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Sample records for steel pressure vessel

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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.)

  6. 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.

  7. Irradiation embrittlement of pressure vessel steels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Vacek, M.

    1975-01-01

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

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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.

  13. 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)

  14. 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.)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. Problems in development of pressure vessel steels

    International Nuclear Information System (INIS)

    McMahon, C.Y.

    1980-01-01

    The tendency of steels to intercrystalline fracture at low stresses is the main factor, limiting fracture resistance of steels in agressive media at conventional and elevated temperatures. The reasons for the phenomenon are analyzed. In particular, the role of grain boundary segregations of non-metallic impurities is pointed out. The ways of the problem solving both at the expense of corresponding microstructure control and by means of selection of the steel chemical composition are considered

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

    Science.gov (United States)

    Jansto, Steven G.

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

  2. 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)

  3. 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)

  4. 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)

  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. 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.

  7. 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.

  8. 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

  9. Corrosion fatigue of pressure vessel steels in PWR environments--influence of steel sulfur content

    International Nuclear Information System (INIS)

    Scott, P.M.; Druce, S.G.; Truswell, A.E.

    1984-01-01

    Large effects of simulated light water reactor environments at 288 C on fatigue crack growth in low alloy pressure vessel steels are observed only when specific mechanical, metallurgical, and electrochemical conditions are satisfied simultaneously. In this paper, the relative importance of three key variables--steel impurity content, water chemistry, and flow rate--and their interaction with loading rate or strain rate are examined. In particular, the results of a systematic examination of the influence of a steel's sulfur content are described

  10. Creep of A508/533 Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Richard Wright

    2014-08-01

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

  11. 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.)

  12. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Bertram, W.

    1975-01-01

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  13. 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.

  14. Radiation embrittlement of WWER 440 pressure vessel steel and of some improved steels by western producers

    International Nuclear Information System (INIS)

    Koutsky, J.; Vacek, M.; Stoces, B.; Pav, T.; Otruba, J.; Novosad, P.; Brumovsky, M.

    1982-01-01

    The resistance was studied of Cr-Mo-V type steel 15Kh2MFA to radiation embrittlement at an irradiation temperature of around 288 degC. Studied was the steel used for the manufacture of the pressure vessel of the Paks nuclear reactor in Hungary. The obtained results of radiation embrittlement and hardening of steel 15Kh2MFA were compared with similar values of Mn-Ni-Mo type steels A 533-B and A 508 manufactured by leading western manufacturers within the international research programme coordinated by the IAEA. It was found that the resistance of steel 15Kh2MFA to radiation embrittlement is comparable with steels A 533-B and A 508 by western manufacturers. (author)

  15. 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

  16. Corrosion fatigue of pressure vessel steels in PWR environments--influence of steel sulfur content

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.M.; Druce, S.G.; Truswell, A.E.

    1984-07-01

    Large effects of simulated light water reactor environments at 288 C on fatigue crack growth in low alloy pressure vessel steels are observed only when specific mechanical, metallurgical, and electrochemical conditions are satisfied simultaneously. In this paper, the relative importance of three key variables--steel impurity content, water chemistry, and flow rate--and their interaction with loading rate or strain rate are examined. In particular, the results of a systematic examination of the influence of a steel's sulfur content are described.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  19. Microscopic examination of crack growth in a pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Isacsson, M.; Narstroem, T. [Royal Inst. of Tech., Stockholm (Sweden)

    1997-01-01

    A fairly systematic microscopic study concerning ductile and ductile-brittle crack growth in the A508B pressure vessel steel has been performed. The main method of investigation was to subject fracture mechanics specimens (sub-sized three point bend specimens) to predetermined load levels corresponding to different amounts of ductile crack extension. The specimens were then cut perpendicularly to the plane of the crack and the area in front of the crack was examined in a SEM. The object of these examinations was to determine if newly encountered computational results could be correlated to crack extension characteristics and to study whether the mechanism of ductile growth was of the void growth type or of the fast shear mechanism. This is important for further numerical modelling of the process. Both the original material and a specially heat treated piece were investigated. The heat treatment was performed in order to raise the transition temperature to about 60 deg C with the object to provide a more convenient testing situation. Charpy V tests were performed for the specially heat treated material to obtain the temperature dependence of the toughness. This was also studied by performing fracture toughness determination on the same type of specimens as were used for the microscopic study. The heat treatment used fulfilled the above purpose and the microscopic studies provide a good understanding of the micro mechanisms operating in the ductile fracture process for this material. 19 refs, 8 figs, 3 tabs.

  20. Microscopic examination of crack growth in a pressure vessel steel

    International Nuclear Information System (INIS)

    Isacsson, M.; Narstroem, T.

    1997-01-01

    A fairly systematic microscopic study concerning ductile and ductile-brittle crack growth in the A508B pressure vessel steel has been performed. The main method of investigation was to subject fracture mechanics specimens (sub-sized three point bend specimens) to predetermined load levels corresponding to different amounts of ductile crack extension. The specimens were then cut perpendicularly to the plane of the crack and the area in front of the crack was examined in a SEM. The object of these examinations was to determine if newly encountered computational results could be correlated to crack extension characteristics and to study whether the mechanism of ductile growth was of the void growth type or of the fast shear mechanism. This is important for further numerical modelling of the process. Both the original material and a specially heat treated piece were investigated. The heat treatment was performed in order to raise the transition temperature to about 60 deg C with the object to provide a more convenient testing situation. Charpy V tests were performed for the specially heat treated material to obtain the temperature dependence of the toughness. This was also studied by performing fracture toughness determination on the same type of specimens as were used for the microscopic study. The heat treatment used fulfilled the above purpose and the microscopic studies provide a good understanding of the micro mechanisms operating in the ductile fracture process for this material

  1. Kinetics of annealing of irradiated surveillance pressure vessel steel

    International Nuclear Information System (INIS)

    Harvey, D.J.; Wechsler, M.S.

    1982-01-01

    Indentation hardness measurements as a function of annealing were made on broken halves of Charpy impact surveillance samples. The samples had been irradiated in commercial power reactors to a neutron fluence of approximately 1 x 10 18 neutrons per cm 2 , E > 1 MeV, at a temperature of about 300 0 C (570 0 F). Results are reported for the weld metal, which showed greater radiation hardening than the base plate or heat-affected zone material. Isochronal and isothermal anneals were conducted on the irradiated surveillance samples and on unirradiated control samples. No hardness changes upon annealing occurred for the control samples. The recovery in hardness for the irradiated samples took place mostly between 400 and 500 0 C. Based on the Meechan-Brinkman method of analysis, the activation energy for annealing was found to be 0.60 +- 0.06 eV. According to computer simulation calculations of Beeler, the activation energy for migration of vacancies in alpha iron is about 0.67 eV. Therefore, the results of this preliminary study appear to be consistent with a mechanism of annealing of radiation damage in pressure vessel steels based on the migration of radiation-produced lattice vacancies

  2. Research to sustain cases for Magnox-reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Graham, W.J.

    1997-01-01

    Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

  3. Residual stresses in a weldment of pressure vessel steel

    International Nuclear Information System (INIS)

    Gott, K.E.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Chi, Byung Ha; Kim, Jeong Tae

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

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

  6. The electrogas and electroslag multipass high speed welding of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Eichhorn, F.; Hirsch, P.; Langenbahn, H.W.; Wubbels, B.

    1978-01-01

    High-speed electroslag and electrogas welding of 15 Mn Ni63 steel plates to achieve high strength and toughness joints for reactor pressure vessels are described. Mechanical testing of overheating-resistant, brittle fracture resistant low alloy steels is discussed. (UK)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-15

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

  10. 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.)

  11. The reactor vessel steels

    International Nuclear Information System (INIS)

    Bilous, W.; Hajewska, E.; Szteke, W.; Przyborska, M.; Wasiak, J.; Wieczorkowski, M.

    2005-01-01

    In the paper the fundamental steels using in the construction of pressure vessel water reactor are discussed. The properties of these steels as well as the influence of neutron irradiation on its degradation in the time of exploitation are also done. (authors)

  12. 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)

  13. Study of radiation damage of steels for light water pressure vessels at UJV

    International Nuclear Information System (INIS)

    Vacek, N.; Stoces, B.

    1980-01-01

    Preoperational determination of radiation resistance of pressure vessel steels is performed at accelerated neutron exposure in a test or materials research reactor. The results obtained at accelerated and operating exposure are not fully identical and surveillance bodies are therefore used manufactured from the pressure vessel material. Currently, the following steels are used for the manufacture of light water reactor pressure vessels: Mn-Mo-Ni (ASTM-A533-B, ASTM-A508), Cr-Mo-V (15Kh2M1FA). At UJV Rez, for irradiation Chanca-M probes imported from France are used featuring electric temperature control. Almost identical radiation embrittlement was measured for all three steels after irradiation with a neutron fluence of 3x10 23 n.m -2 at a temperature of 290 degC. (H.S.)

  14. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology program series 4 and 5)

    International Nuclear Information System (INIS)

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

    1985-01-01

    This report presents studies on the irradiation effects in low-alloy reactor pressure vessel steels. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (''current practice welds''). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds. 27 refs., 22 figs

  15. Estimation of residual stresses in reactor pressure vessel steel specimens clad by stainless steel strip electrodes

    International Nuclear Information System (INIS)

    Schimmoeller, H.A.; Ruge, J.L.

    1978-01-01

    The equations to determine a two-dimensional state of residual stress in flat laminated plates are well known from an earlier work by one of the authors. The derivation of these equations leads to a linear, inhomogeneous system of Volterra's integral equations of the second kind. To ascertain the unknown residual stresses from these equations it is necessary to cut down the thickness of the test plate layer by layer. This results in two-dimensional deformation reactions in the rest of the test plate, which can be measured, e.g. by a strain gauge rosette applied to the opposite side of the plate. The above-mentioned stress analysis has been transferred to 86mm thick reactor pressure vessel steel specimens (Type 22NiMoCr 37, DIN-No. 1.6751, similar to ASTM A508, Class 2) double-run clad by austenitic stainless steel strip electrodes (first layer 24/13 Cr-Ni steel, second layer 21/10 Cr-Ni steel). The overall dimensions of the clad specimens investigated amounted to 200 x 200 x (86+4.5+4.5)mm. At the surface of the austenitic cladding there is a two-dimensional tensile normal stress state of about 200N/mm 2 parallel, and about 300N/mm 2 transverse, to the welding direction. The maximum tensile stress was 8mm below the interface (fusion line, material transition) in the parent material. The stress distributions of the specimens investigated, determined on the basis of the above-mentioned combined experimental mathematical procedure, are presented graphically for the as-welded (as-delivered) and annealed (600 0 C/12hr) conditions. (author)

  16. Effect of stress relief parameters on the mechanical properties of pressure vessel steels and weldments

    International Nuclear Information System (INIS)

    Canonico, D.A.; Stelzman, W.J.

    1976-01-01

    Post weld heat treatments of thick-section A533B steel for nuclear pressure vessels are discussed with reference to the ASME code. The discussion is in the form of a lecture and summarized by noting that the ASME code, in particular Section III, Division 1, imposes a post weld heat treatment requirement on pressure vessels fabricated from low alloy high strength steels. The Code permits a holding temperature range, the high side of which could result in poorer toughness properties. Long times in excess of 100 hours and/or high temperatures, 649 0 C can result in an increase in the NDT and a decrease in the upper shelf energy

  17. Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

    International Nuclear Information System (INIS)

    James, L.A.

    1977-01-01

    The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  19. Updated embrittlement trend curve for reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Ikonen, K.

    2001-01-01

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

  2. Study of cladding toughness in a pressure vessel steel water reactor

    International Nuclear Information System (INIS)

    Soulat, P.; Al Mundheri, M.

    1984-12-01

    Toughness of cladding and pressure vessel steel were determined at different temperatures in order to appreciate the participation of cladding resistance against crack propagation. The toughness of cladding is comparable with typical results on austenitic welds. The test on covered CT specimens shows the possibility of having a relatively good prevision of the behaviour of a coated structure

  3. Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1990-11-01

    Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Joining dissimilar stainless steels for pressure vessel components

    International Nuclear Information System (INIS)

    Zheng Sun; Huai-Yue Han

    1994-01-01

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

  6. Hydrogen induced plastic damage in pressure vessel steel of 2.25Cr-1Mo

    International Nuclear Information System (INIS)

    Han, G.W.; Song, Y.J.

    1995-01-01

    2.25Cr-1Mo steel is generally employed as a hydrogenation reaction vessel material used at elevated temperature and in a hydrogen containing environment. During service of the reaction vessel, a large number of hydrogen atoms would enter its wall. When the reaction vessel is shutdown and the temperature reduces to about ambient temperature, the hydrogen atoms remaining in the wall would induce plastic damage in the steel. The mechanism of hydrogen induced plastic damage is different for various materials with different microstructures. Investigations have demonstrated that the hydrogen induced plastic damage in carbide annealed carbon steels is caused by hydrogen accelerating the initiating and growing of microvoids from the carbide particles. However, SEM examination on the fracture surface of hydrogen charged tensile specimen of 2.25Cr-1Mo steel show that a large number of fisheyes appear on the fracture surface. This indicates that hydrogen induced plastic damage in 2.25Cr-1Mo steel is related to the occurrence of fisheye cracks during plastic deformation. By means of micro-fracture mechanics to analyze fisheye crack occurrence from the first generation microvoid, the mechanism of hydrogen induced plastic damage in the pressure vessel steel is investigated

  7. Influence of heat treatments on thermoelectric power of pressure vessel steels: effect of microstructural evolutions of strongly segregated areas

    International Nuclear Information System (INIS)

    Houze, M.

    2002-12-01

    Thermoelectric power measurement (TEP) is a very potential non destructive evaluation method considered to follow ageing under neutron irradiation of pressure vessel steel of nuclear reactor. Prior to these problems, the aim of this study is to establish correlations between TEP variations and microstructural evolutions of pressure vessel steels during heat treatments. Different steels, permitting to simulate heterogeneities of pressure vessel steels and to deconvoluate main metallurgical phenomenons effects were studied. This work allowed to emphasize effect on TEP of: austenitizing and cooling conditions and therefore of microstructure, metallurgical transformations during tempering (recovery, precipitation of alloying elements), and particularly molybdenum precipitation associated to secondary hardening, residual austenite amount or partial austenitizing. (author)

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

    International Nuclear Information System (INIS)

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

    1976-12-01

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

  10. Pressure vessels for reactors made from structural steel with limited tensile strength

    International Nuclear Information System (INIS)

    Machatti, H.

    1973-01-01

    The reactor pressure vessel is prestressed in several directions with prestressing elements fabricated of steel with a high yielding point. This design allows a substantial reduction of wall thickness or an increase of the inner diameter at equal wall thickness. The prestress of the prestressing elements is designed to achieve a maximum stress release of the vessel walls at normal operating conditions and to fully utilize the maximum load of the vessel walls. For safety reasons the cross section of the prestressing elements is constructed in a way that strain is always 20 % lower the yield point. (P.K.)

  11. Hydrogen attack of pressure-vessel steel. Progress report, April 1, 1980-March 31, 1981

    International Nuclear Information System (INIS)

    Shewmon, P.G.

    1980-12-01

    The nucleation and growth of methane bubbles in the hydrogen attack of pressure vessel steel has been shown to obey models developed to describe the growth of bubbles limiting the creep ductility of metals. This has been done through studies of the effect of prior deformation on bubble nucleation as well as the effect of methane pressure (stress) and temperature on growth kinetics. A comprehensive model of the factors limiting growth has been developed. Its application to the hydrogen attack of a 2 1/4 Cr-1 Mo steel leads to several interesting predictions

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

    International Nuclear Information System (INIS)

    1998-10-01

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

  13. APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Blavette, D.

    1996-01-01

    Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

  14. Irradiation effects in strain aged pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M; Myers, H P

    1962-02-15

    Tensile specimens, Charpy-V notch and subsize impact specimens of an aluminium killed carbon manganese steel, have been irradiated at 160 - 190 deg C in the reactor G1. The total neutron dose received was 2.4 x 10{sup 18} n/cm{sup 2} (> 1 MeV). Specimens were prepared from normalized plate and from strain aged material from the same plate. It was found that the changes in brittle ductile transition temperature due to neutron irradiation and those due to strain ageing must be considered additive.

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

    International Nuclear Information System (INIS)

    Tice, D.R.

    1987-01-01

    1) Since environmentally assisted cracking (EAC) is a time dependent process, assessment should be based on time rather than cycle dependent parameters. Thus an a/sub e/ vs a/sub i/ (or strain rate) basis for assessment should be used in preference to da/dN vs ΔK. 2) The threshold strain rate or velocity for the onset of EAC is controlled by material and environmental factors (e.g. steel sulphur content and water chemistry), and possibly by mechanical loading factors such as R ratio and load interaction effects. Above the threshold, crack growth rates are usually unacceptably rapid. 3) Sample calculations show that predicted crack growth rates using a time based model can be below or above those calculated using ASME XI depending on the value of the EAC threshold velocity but that for normal PWR operating conditions rates are likely to be below those predicted by the ASME code

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  18. Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

  19. 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

  20. Apparent embrittlement saturation and radiation mechanisms of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Pachur, D.

    1981-01-01

    The irradiation and annealing results of three different reactor pressure vessel steels are reported. Steel A, a basic material according to ASTM A-533 B having 0.15 percent vanadium; and Steel C contained 3.2 percent nickel. The steels were irradiated at 150, 300, and 400 degree C with neutron fluxes of 6 multiplied by 10 11 and 3 multiplied by 10 13 neutrons (n)/cm 2 /s. An apparent saturation-in-irradiation effect was found within certain neutron fluence ranges. During the annealing, various recovery processes occur in different temperature ranges. These are characterized by various activation energies. The individual processes were determined by the different time dependencies at various temperatures. Two causes for the apparent saturation were discovered from the behavior of the annealing curves

  1. Reactor pressure vessel steels ASTM A533B and A508 Cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Kemppainen, M.; Toerroenen, K.

    1979-11-01

    This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  3. Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

    International Nuclear Information System (INIS)

    Klaar, H.J.

    1975-01-01

    As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

  4. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.

    1978-01-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs

  5. Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Wu, Sujun; Jin, Huijin; Sun, Yanbin; Cao, Luowei

    2014-01-01

    The critical cleavage fracture stress of SA508 Gr.4N and SA508 Gr.3 low alloy reactor pressure vessel (RPV) steels was studied through the combination of experiments and finite element method (FEM) analysis. The results showed that the value of the local cleavage fracture stress, σ F , of SA508 Gr.4N steel was significantly higher than that of SA508 Gr.3 steel. Detailed microstructural analysis was carried out using FEGSEM which revealed much smaller grains, finer and more homogenous carbide particles formed in SA508 Gr.4N steel. Compared with the SA508 Gr.3 steel currently used in the nuclear industry, the SA508 Gr.4N steel possesses higher strength and notch toughness as well as improved cleavage fracture behavior, and is considered a better candidate RPV steel for the next generation nuclear reactors. - Highlights: • Critical cleavage fracture stress was calculated through experiments and FEM. • Effects of both grain and carbide particle sizes on σ F were discussed. • The SA508 Gr.4N steel is a better candidate for the next generation nuclear reactors

  6. Relationships between Charpy impact shelf energies and upper shelf Ksub(IC) values for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Witt, F.J.

    1983-01-01

    Charpy shelf data and lower bound estimates of Ksub(IC) shelf data for the same steels and test temperatures are given. Included are some typical reactor pressure vessel steels as well as some less tough or degraded steels. The data were evaluated with shelf estimates of Ksub(IC) up to and exceeding 550 MPa√m. It is shown that the high shelf fracture toughness representative of tough reactor pressure vessel steels may be obtained from a knowledge of the Charpy shelf energies. The toughness transition may be obtained either by testing small fracture toughness specimens or by Charpy energy indexing. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  9. Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

    International Nuclear Information System (INIS)

    Ghoneim, M.M.

    1987-05-01

    The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

  10. The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

    Energy Technology Data Exchange (ETDEWEB)

    Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

    1994-12-31

    Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

  11. Acoustic emission during the elastic-plastic deformation of low alloy reactor pressure vessel steels. I

    International Nuclear Information System (INIS)

    Holt, J.; Goddard, D.J.

    1980-01-01

    Measurements of the acoustic emission behaviour of A533B and C-Mn low alloy reactor pressure vessel steels subjected to uniaxial tensile deformation are described. The effects on the emission activity of the rolling plane orientation and the carbide morphology were examined. Detailed discussions are given of the stress dependence of the emission activity below yield and of its recovery by annealing at the stress relief temperature. It is shown that the dominant emission source is the same in both steels and is associated with inclusions, such as MnS, elongated by the rolling process, the carbide morphology being relatively unimportant. A criterion for the occurrence of an emission is obtained which is directly analogous to the general criterion for yielding. It is also shown that a large fraction, at least, of the emission activity arises from a recoverable process such as localized yielding around inclusions or limited inclusion decohesion and not from inclusion fracture. Low activity in C-Mn steel taken from reactor pressure vessels, previously attributed to spheroidization of carbides, is shown to be due to the limited acoustic recovery of these relatively high sulphur content steels when annealed at the stress relief temperature. It is concluded that the limited amplitudes of these emissions during deformation severely restrict their potential application in practice. (Auth.)

  12. Effect of heterogeneities on the thermoelectric power of pressure vessel steel

    International Nuclear Information System (INIS)

    Simonet, L.

    2006-12-01

    In service working conditions, the vessel of the Pressurized Water Reactors (PWR) undergoes an ageing due to irradiation. In order to follow the evolution of the mechanical characteristics of the steel in service, EDF launched a surveillance program which consists to carry out mechanical tests on samples aged in reactor. However, the results of these tests have the disadvantage to be affected by the presence of heterogeneities within the steel. Indeed, because of its manufacturing process, the steel contains segregated areas. Thus, EDF launched Thermoelectric Power Measurements (TEP) on the resilience samples of the surveillance program, to complete the mechanical tests and to help with their interpretation. However, these measurements are today difficult to analyse because they include at the same time the effect of the irradiation and the effect of the metallurgical heterogeneities. The aim of this work consisted in evaluating the effect of the heterogeneities on the TEP of the non-irradiated vessel steel. For that, a numerical model was developed which allows to calculate the TEP of a composite structure. We have shown that the model is pertinent to highlight the effect of the heterogeneities on the TEP of the vessel steel, which is considered like a 'matrix'/'segregation' composite. The model allowed us to put emphasis on the influence of different parameters on the TEP measurement. We have thus showed that the measurements conditions have an important effect on the obtained TEP value (influence of the applied pressure, the position of the sample on the device, the site of the metallurgical heterogeneities,...). (author)

  13. Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

    International Nuclear Information System (INIS)

    Tipping, P.; Ineichen, U.; Cripps, R.

    1994-01-01

    The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Fracture toughness and crack growth resistance of pressure vessel plate and weld metal steels

    International Nuclear Information System (INIS)

    Moskovic, R.

    1988-01-01

    Compact tension specimens were used to measure the initiation fracture toughness and crack growth resistance of pressure vessel steel plates and submerged arc weld metal. Plate test specimens were manufactured from four different casts of steel comprising: aluminium killed C-Mn-Mo-Cu and C-Mn steel and two silicon killed C-Mn steels. Unionmelt No. 2 weld metal test specimens were extracted from welds of double V butt geometry having either the C-Mn-Mo-Cu steel (three weld joints) or one particular silicon killed C-Mn steel (two weld joints) as parent plate. A multiple specimen test technique was used to obtain crack growth data which were analysed by simple linear regression to determine the crack growth resistance lines and to derive the initiation fracture toughness values for each test temperature. These regression lines were highly scattered with respect to temperature and it was very difficult to determine precisely the temperature dependence of the initiation fracture toughness and crack growth resistance. The data were re-analysed, using a multiple linear regression method, to obtain a relationship between the materials' crack growth resistance and toughness, and the principal independent variables (temperature, crack growth, weld joint code and strain ageing). (author)

  16. Effects of temperature on corrosion fatigue crack growth of pressure vessel steels in PWR coolant

    International Nuclear Information System (INIS)

    Tice, D.R.; Bramwell, I.L.; Fairbrother, H.; Worswick, D.

    1994-01-01

    This paper presents experimental results concerning crack propagation rates in A508-III pressure vessel steel (medium sulphur content) exposed to PWR primary water at temperatures between 130 and 290 C. The results indicate that the greatest increase in corrosion fatigue crack growth rate occurs at temperatures in the range 150 to 200 C. Under these conditions, there was a marked change in the appearance of the fracture surface, with extensive micro-branching of the crack front and occasional bifurcation of the whole crack path. In contrast, at 290 C, the fracture surface is smoother, similar to that due to inert fatigue. The implication of these observations for assessment of the pressure vessel integrity, is examined. 14 refs., 15 figs., 3 tabs

  17. Application of small specimens to fracture mechanics characterization of irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Sokolov, M.A.; Wallin, K.; McCabe, D.E.

    1996-01-01

    In this study, precracked Charpy V-notch (PCVN) specimens were used to characterize the fracture toughness of unirradiated and irradiated reactor pressure vessel steels in the transition region by means of three-point static bending. Fracture toughness at cleavage instability was calculated in terms of elastic-plastic K Jc values. A statistical size correction based upon weakest-link theory was performed. The concept of a master curve was applied to analyze fracture toughness properties. Initially, size-corrected PCVN data from A 533 grade B steel, designated HSST Plate O2, were used to position the master curve and a 5% tolerance bound for K Jc data. By converting PCVN data to IT compact specimen equivalent K Jc data, the same master curve and 5% tolerance bound curve were plotted against the Electric Power Research Institute valid linear-elastic K Jc database and the ASME lower bound K Ic curve. Comparison shows that the master curve positioned by testing several PCVN specimens describes very well the massive fracture toughness database of large specimens. These results give strong support to the validity of K Jc with respect to K Ic in general and to the applicability of PCVN specimens to measure fracture toughness of reactor vessel steels in particular. Finally, irradiated PCVN specimens of other materials were tested, and the results are compared to compact specimen data. The current results show that PCVNs demonstrate very good capacity for fracture toughness characterization of reactor pressure vessel steels. It provides an opportunity for direct measurement of fracture toughness of irradiated materials by means of precracking and testing Charpy specimens from surveillance capsules. However, size limits based on constraint theory restrict the operational test temperature range for K Jc data from PCVN specimens. 13 refs., 8 figs., 1 tab

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

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

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

  19. Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

  20. Influence of steel-making process and heat-treatment temperature on the fatigue and fracture properties of pressure vessel steels

    International Nuclear Information System (INIS)

    Koh, S. K.; Na, E. G.; Baek, T. H.; Won, S. Y.; Park, S. J.; Lee, S. W.

    2001-01-01

    In this paper, high strength pressure vessel steels having the same chemical compositions were manufactured by the two different steel-making processes, such as Vacuum Degassing(VD) and Electro-Slag Remelting(ESR) methods. After the steel-making process, they were normalized at 955 deg. C, quenched at 843 .deg. C, and finally tempered at 550 .deg. C or 450 deg. C, resulting in tempered martensitic microstructures with different yielding strengths depending on the tempering conditions. Low-Cycle Fatigue(LCF) tests, Fatigue Crack Growth Rate(FCGR) tests, and fracture toughness tests were performed to investigate the fatigue and fracture behaviors of the pressure vessel steels. In contrast to very similar monotonic, LCF, and FCGR behaviors between VD and ESR steels, a quite difference was noticed in the fracture toughness. Fracture toughness of ESR steel was higher than that of VD steel, being attributed to the removal of impurities in steel-making process

  1. Underwater cutting of stainless steel plate and pipe for dismantling reactor pressure vessels

    International Nuclear Information System (INIS)

    Hamasaki, M.; Tateiwa, F.; Kanatani, F.; Yamashita, S.

    1982-01-01

    A consumable electrode water jet cutting technique is described. Satisfactory underwater cutting of 80mm stainless steel plate using a current of 2000A and at a water depth of 200mm has been demonstrated. The electrical requirements for this arc welding method applied to cutting were found to be approximately one third those required for conventional plasma arc cutting for the same thickness plate. An application of this technique might be found in the dismantling of atomic reactor pressure vessels, and parts of commercial atomic reactors. (author)

  2. Temperature dependence of the fracture toughness and the cleavage fracture strength of a pressure vessel steel

    International Nuclear Information System (INIS)

    Kotilainen, H.

    1980-01-01

    A new model for the temperature dependence of the fracture toughness has been sought. It is based on the yielding processes at the crack tip, which are thought to be competitive with fracture. Using this method a good correlation between measured and calculated values of fracture toughness has been found for a Cr-Mo-V pressure vessel steel as well as for A533B. It has been thought that the application of this method can reduce the number of surveillance specimens in nuclear reactors. A method for the determination of the cleavage fracture strength has been proposed. 28 refs

  3. Progress in Investigation of WWER-440 Reactor Pressure Vessel Steel by Gamma and Moessbauer Spectroscopy

    International Nuclear Information System (INIS)

    Hascik, J.; Slugen, V.; Lipka, J.; Hinca, R.; Toth, I.; Groene, R.; Uvacik, P.; Kupca, L.

    1998-01-01

    Gamma spectroscopic analyse and first experimental results of original irradiated reactor pressure vessel surveillance specimens are discussed in. In 1994, the new ''Extended Surveillance Specimen Program for nuclear Reactor Material Study'' was started in collaboration with the nuclear power plants (NPP) V-2 Bohunice (Slovakia). The first batch of MS samples (after 1 year, which is equivalent to 5 years of loading RPV-steel) was measured and interpreted using the new four components approach with the aim to observe microstructural changes due to thermal and neutron treatment resulting from operating conditions in NPP. The systematic changes in the relative areas of Moessbauer spectra components were observed. (author)

  4. Strain ageing of nuclear pressure vessel steels A533B and A508 cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Toerroenen, K.

    1978-04-01

    The susceptibility of the reactor pressure vessel steels A533B and A508 cl.2 to strain ageing has been studied using conventional tensile and impact testing of prestrained and aged specimens. The results show a modest susceptibility, seen as an increase in yield strength and Charpy V transition temperatures. The effect of varying alloying additions within the range of normal production was not observed, but the initial mechanical properties clearly affect the strain ageing. The lower the initial yield strength, the higher increase in strength and the lower increase in transition temperature is observed. (author)

  5. Revision of the fracture models in steels for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Darwish, F A.I. [Pontificia Univ. Catolica do Rio de Janeiro (Brazil). Dept. de Ciencia dos Materiais e Metalurgia

    1981-01-01

    The variation of toughness with the temperature of steels used in the fabrication of nuclear pressure vessels is presented and discuted by mathematical models aiming to reach a critical value of stress or deformation at the moment of the fracture. The mathematical model considered are compatible with the fracture micromechanisms in action and they are capable of foreseeing the variations in the toughness from the mechanical properties evaluated in the tension test. The neutron irradiation effects in the toughness as well as in the variation of this toughness with the operating temperature are still described.

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

    International Nuclear Information System (INIS)

    McHenry, H.I.; Alers, G.A.

    1998-01-01

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

  7. 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

  8. Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

    1993-01-01

    In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

  9. Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ``clouds`` more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs.

  10. The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

    International Nuclear Information System (INIS)

    Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

    2009-01-01

    This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

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

    CERN Document Server

    Bouchard, P J; Treimer, W

    2002-01-01

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

  12. Characterization of atom clusters in irradiated pressure vessel steels and model alloys

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

    1993-12-01

    In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  14. Guide to the periodic inspection of nuclear reactor steel pressure vessels

    International Nuclear Information System (INIS)

    1969-01-01

    This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-31

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

  16. Application of electron beam welding to large size pressure vessels made of thick low alloy steel

    International Nuclear Information System (INIS)

    Kuri, S.; Yamamoto, M.; Aoki, S.; Kimura, M.; Nayama, M.; Takano, G.

    1993-01-01

    The authors describe the results of studies for application of the electron beam welding to the large size pressure vessels made of thick low alloy steel (ASME A533 Gr.B cl.2 and A533 Gr.A cl.1). Two major problems for applying the EBW, the poor toughness of weld metal and the equipment to weld huge pressure vessels are focused on. For the first problem, the effects of Ni content of weld metal, welding conditions and post weld heat treatment are investigated. For the second problem, an applicability of the local vacuum EBW to a large size pressure vessel made of thick plate is qualified by the construction of a 120 mm thick, 2350 mm outside diameter cylindrical model. The model was electron beam welded using local vacuum chamber and the performance of the weld joint is investigated. Based on these results, the electron beam welding has been applied to the production of a steam generator for a PWR. (author). 3 refs., 10 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  18. Models for ductile crack initiation and tearing resistance under mode 1 loading in pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, M.R.

    1988-06-01

    Micromechanistic models are presented which aim to predict plane strain ductile initiation toughness, tearing resistance and notched bar fracture strains in pressure vessel steels under monotonically increasing tensile (mode 1) loading. The models for initiation toughness and tearing resistance recognize that ductile fracture proceeds by the growth and linkage of voids with the crack-tip. The models are shown to predict the trend of initiation toughness with inclusion spacing/size ratio and can bound the available experimental data. The model for crack growth can reproduce the tearing resistance of a pressure vessel steel up to and just beyond crack growth initiation. The fracture strains of notched bars pulled in tension are shown to correspond to the achievement of a critical volume fraction of voids. This criterion is combined with the true stress - true strain history of a material point ahead of a blunting crack-tip to predict the initiation toughness. An attempt was made to predict the fracture strains of notched tensile bars by adopting a model which predicts the onset of a shear localization phenomenon. Fracture strains of the correct order are computed only if a ''secondary'' void nucleation event at carbide precipitates is taken into account. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Boåsen, Magnus, E-mail: boasen@kth.se [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Ehrnstén, Ulla [VTT Technical Research Centre of Finland Ltd, PO Box 1000, FI-02044 VTT (Finland)

    2017-02-15

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. - Highlights: • Hardness testing is combined with post irradiation annealing at 330, 360 and 390 °C. • Unstable matrix defects is studied in a reactor pressure vessel steel. • Comparison between surveillance material and accelerated irradiation. • No evidence of unstable matrix defects, i.e. not present in studied material. • Difference in hardness recovery between irradiation conditions found at 390 °C.

  20. Failure prediction of low-carbon steel pressure vessel and cylindrical models

    International Nuclear Information System (INIS)

    Zhang, K.D.; Wang, W.

    1987-01-01

    The failure loads predicted by failure assessment methods (namely the net-section stress criterion; the EPRI engineering approach for elastic-plastic analysis; the CEGB failure assessment route; the modified R6 curve by Milne for strain hardening; and the failure assessment curve based on J estimation by Ainsworth) have been compared with burst test results on externally, axially sharp notched pressure vessel and open-ended cylinder models made from typical low-carbon steel St45 seamless tube which has a transverse true stress-strain curve of straight-line and parabola type and a high value of ultimate strength to yield. It was concluded from the comparison that whilst the net-section stress criterion and the CEGB route did not give conservative predictions, Milne's modified curve did give a conservative and good prediction; Ainsworth's curve gave a fairly conservative prediction; and EPRI solutions also could conditionally give a good prediction but the conditions are still somewhat uncertain. It is suggested that Milne's modified R6 curve is used in failure assessment of low-carbon steel pressure vessels. (author)

  1. Stress corrosion cracking studies of reactor pressure vessel steels. Final report

    International Nuclear Information System (INIS)

    Van Der Sluys, W.A.

    1996-10-01

    The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

  2. Re-austenitisation of chromium-bearing pressure vessel steels during the weld thermal cycle

    International Nuclear Information System (INIS)

    Dunne, Druce; Li, Huijun; Jones, Christopher

    2013-01-01

    Steels with chromium contents between 0.5 and 12 wt% are commonly used for fabrication of creep resistant pressure vessels (PV) for the power generation industry. Most of these steels are susceptible to Type IV creep failure in the intercritical and/ or grain refined regions of the heat affected zone (HAZ) of the parent metal. The re-austenitisation process plays a central role in establishing the transformed microstructures and the creep resistance of the various sub-zones of the HAZ. The high alloy content and the presence of alloy-rich carbides in the as-supplied parent plate can significantly retard the kinetics of transformation to austenite, resulting in both incomplete austenitisation and inhomogeneous austenite. Overlapping weld thermal cycles in multi-pass welds add further complexity to the progressive development of microstructure over the course of the welding process. In order to clarify structural evolution, thermal simulation has been used to study the effects of successive thermal cycles on the structures and properties of the HAZ of 2.25Cr-1Mo steel. The results showed that, before post-weld heat treatment (PWHT), the HAZ microstructures and properties, particularly in doubly reheated sub-zones, were highly heterogeneous and differed markedly from those of the base steel. It is concluded that close control of the thermal cycle by pre-heat, weld heat input and post-heat is necessary to obtain a heat affected zone with microstructures and properties compatible with those of the base plate.

  3. Evolution of manganese–nickel–silicon-dominated phases in highly irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Wells, Peter B.; Yamamoto, Takuya; Miller, Brandon; Milot, Tim; Cole, James; Wu, Yuan; Odette, G. Robert

    2014-01-01

    Formation of a high density of Mn–Ni–Si nanoscale precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels could lead to severe unexpected embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement prediction models, would emerge only at high fluence. However, the mechanisms and variables that control Mn–Ni–Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni contents were carried out at ∼295 °C to high and very high neutron fluences of ∼1.3 × 10 20 and ∼1.1 × 10 21 n cm −2 . Atom probe tomography shows that significant mole fractions of Mn–Ni–Si-dominated precipitates form in the Cu-bearing steels at ∼1.3 × 10 20 n cm −2 , while they are only beginning to develop in Cu-free steels. However, large mole fractions of these precipitates, far in excess of those found in previous studies, are observed at 1.1 × 10 21 n cm −2 at all Cu contents. At the highest fluence, the precipitate mole fractions primarily depend on the alloy Ni, rather than Cu, content. The Mn–Ni–Si precipitates lead to very large increases in measured hardness, corresponding to yield strength elevations of up to almost 700 MPa

  4. Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Kristina, E-mail: kristina.lindgren@chalmers.se [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Boåsen, Magnus [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Stiller, Krystyna [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Vattenfall Ringhals AB, SE-430 22 Väröbacka (Sweden); Thuvander, Mattias [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

    2017-05-15

    Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher. - Highlights: •Clustering in a low Cu, high Ni reactor pressure vessel steel weld is studied. •The clusters nucleate and grow during irradiation, and consist of Ni, Mn, Si, and Cu. •High flux neutron irradiated material is compared to surveillance material. •High flux was found to result in smaller clusters with a larger number density. •Hardness follows the same dependence on fluence, independent of flux.

  5. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Pareige, P.

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  7. Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels

    International Nuclear Information System (INIS)

    Lott, R.G.; Freyer, P.D.

    1996-01-01

    The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior

  8. Fatigue of non-welded pressure vessels made of high strength steel

    International Nuclear Information System (INIS)

    Rauscher, F.

    2003-01-01

    When using high strength steels for pressure vessels, cyclic fatigue requirements may become decisive for the design. Within a European research project, two typical non-welded types of vessels--gas cylinders as used for gas transportation and hydraulic accumulators with screwed in ends--were investigated. The results of the fatigue analyses and of the testing of these vessels are described here. Special attention is drawn to the evaluation of the stresses in the threads used for threaded in flat ends and rings, because the usual formulae for bolted connections cannot be used. In the case of sharp notches and of threads, the experiments showed that the fatigue calculation gave conservative results. The unexpected failure of the gas cylinders in the cylindrical part and at the onset of the end showed that the fatigue analyses according to prEN13445-3 clause 18 is non-conservative for these surfaces without mechanical preparation, and need special consideration. Based on the investigations, a stress concentration factor for small fabrication notches and a new surface finish factor is proposed

  9. Fatigue of non-welded pressure vessels made of high strength steel

    Energy Technology Data Exchange (ETDEWEB)

    Rauscher, F

    2003-03-01

    When using high strength steels for pressure vessels, cyclic fatigue requirements may become decisive for the design. Within a European research project, two typical non-welded types of vessels--gas cylinders as used for gas transportation and hydraulic accumulators with screwed in ends--were investigated. The results of the fatigue analyses and of the testing of these vessels are described here. Special attention is drawn to the evaluation of the stresses in the threads used for threaded in flat ends and rings, because the usual formulae for bolted connections cannot be used. In the case of sharp notches and of threads, the experiments showed that the fatigue calculation gave conservative results. The unexpected failure of the gas cylinders in the cylindrical part and at the onset of the end showed that the fatigue analyses according to prEN13445-3 clause 18 is non-conservative for these surfaces without mechanical preparation, and need special consideration. Based on the investigations, a stress concentration factor for small fabrication notches and a new surface finish factor is proposed.

  10. Micromechanisms of ductile stable crack growth in nuclear pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Belcher, W.P.A.; Druce, S.G.

    1981-10-01

    The objective of this work was to investigate the relationship between the micromechanisms of ductile crack growth, the microstructural constituent phases present in nuclear pressure vessel steel, and the observed fracture behavior as determined by impact and fracture mechanics tests. Results from a microstructural and mechanical property comparison of an A508 Class 3 pressurized water reactor nozzle forging cutout and a 150-mm-thick A533B Class 1 plate are reported. The variation of upper-shelf toughness between the two steels and its orientation sensitivity are discussed on the basis of inclusion and precipitate distributions. Inclusion clusters in A533B, deformed to elongated disks in the rolling plane, have a profound effect on short transverse fracture properties. Data derived using the multi-specimen J-integral method to characterize the initiation of ductile crack extension and resistance to stable crack growth are compared with equivalent Charpy results. Results of the J /SUB R/ -curve analyses indicate (1) that the A533B short transverse crack growth resistance is approximately half that observed from transverse and longitudinal specimen orientations, and (2) that the A508 initiation toughness and resistance to stable crack growth are insensitive to position through the forging wall, and are higher than exhibited by A533B at any orientation in the midthickness position.

  11. Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

    International Nuclear Information System (INIS)

    Burke, M.G.; Freyer, P.D.; Mager, T.R.

    1993-01-01

    In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ''precipitation-type'' and a ''damage-type'' component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs

  12. The IAEA data base ageing of reactor pressure vessel steels and welds

    International Nuclear Information System (INIS)

    Gillemot, F.; Ianko, L.; Davies, L.M.

    1995-01-01

    This paper describes one aspect of the International Atomic Energy Agency (IAEA) data base, that is to do with the ageing of reactor pressure vessel (RPV) steels and welds. It describes the background and the need for and the benefits deriving from such an international data base encompassing a greater number of sources than currently incorporated in existing international and national data bases. The paper describes the organization of this data base and the controls necessary for data acquisition and control. The current state of progress is described. Membership of and participation in this project is given and data access is also described. The technical features of the data base are described in terms of the structure of the data base and the hardware and software. New features are proposed such as the inclusion of measured curve data and metallographic data. Technical aspects of data evaluation are also included. (author). 1 ref., 6 figs

  13. Characterization of phosphorus segregation in neutron-irradiated Russian pressure vessel steel weld

    International Nuclear Information System (INIS)

    Miller, M.K.; Jayaram, R.; Russell, K.F.

    1995-01-01

    An atom probe field ion microscopy characterization of three Russian pressure vessel steels has been performed. Field ion micrographs of several lath boundaries have indicated that they are decorated with a semicontinuous film of discrete brightly-imaging precipitates that were identified as molybdenum carbonitrides. In addition, extremely high phosphorus levels were measured at the lath boundaries. The phosphorus was found to be confined to an extremely narrow region indicative of monolayer type segregation. The phosphorus coverage determined from the atom probe results of the unirradiated materials agree with predictions based on McLean's equilibrium model of grain boundary segregation. The boundary phosphorus coverage of a neutron-irradiated weld material was significantly higher than in the unirradiated material. Ultrafine darkly-imaging copper- and phosphorus-enriched precipitates were also observed in the matrix of the neutron-irradiated material. (orig.)

  14. Long-term ageing effects in reactor pressure vessel steels investigated by positron annihilation spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Butterling, Maik; Anwand, Wolfgang; Wagner, Andreas [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Radiation Physics, Dresden (Germany); Bergner, Frank; Ulbricht, Andreas; Wagner, Arne [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Ion Beam Physics and Materials Research, Dresden (Germany)

    2014-07-01

    Neutron irradiation of reactor pressure vessel steels leads to the formation of nano-sized defects which can deteriorate the material. An understanding of the microstructural evolution of the material is important for making reliable security assessments about possible future long-term operation of nuclear power plants. So-called late-blooming phases are formed after long-term irradiation and lead to considerable material ageing effects. Encouraging factors for the formation of these phases are a low Cu-content, moderate to high contents of Mn and Ni, low irradiation temperatures and different neutron fluxes. Positron annihilation lifetime spectroscopy which is ideally suited for the detection and characterization of these irradiation-induced defects was applied for different selected materials which fulfill these conditions in order to investigate the occurrence and behavior of these phases.

  15. Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

    Energy Technology Data Exchange (ETDEWEB)

    Burke, M G; Freyer, P D; Mager, T R

    1994-12-31

    In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ``precipitation-type`` and a ``damage-type`` component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs.

  16. Warm pre-stress experiments on highly irradiated reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Landron, C.; Ait-Bachir, M.; Moinereau, D.; Molinie, E.; Garbay, E.

    2015-01-01

    In the aim to justify in-service integrity of reactor pressure vessel beyond 40 years, experimental warm pre-stress (WPS) tests were performed on irradiated materials representative of RPV steels corresponding to 40 operating years. Different types of WPS loading path have been considered to cover typical postulated accidental transients. These results confirmed the beneficial effect of WPS on the cleavage fracture resistance of the irradiated materials. No fracture occurred during the cooling phase of the loading path and the fracture toughness values are higher than that measured with conventional isothermal tests. The analyses of the experiments, conducted using either simplified engineering models or more refined fracture models based on local approach to cleavage fracture, are in agreement with the experimental results. (authors)

  17. Microstructural parameters and yielding in a quenched and tempered Cr-Mo-V pressure vessel steel

    International Nuclear Information System (INIS)

    Toerroenen, Kari

    1979-01-01

    In this work the plastic deformation behaviour of a Cr-Mo-V pressure vessel steel is studied at ambient and low temperatures. To produce a wide range of microstructures, different austenitizing, quenching and tempering treatments are performed. The microstructures, including grain and dislocation structures as well as carbides, are evaluated. A qualitative model is proposed for the martensitic and bainitic transformations explaining the morphology and crystallography of the transformation products. Based on microstructural observations of undeformed and deformed materials, as well as the tensile test results, the role of various obstacles to dislocation motion in plastic deformation is evaluated. Finally the strength increment, its temperature dependence and the effect due to combinations of various obstacles are analyzed. The results are intended to serve as basis for further fracture behaviour analyses. (author)

  18. The metrological problems of irradiation embrittlement of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Vodenicharov, S.; Kamenova, Ts.

    1993-01-01

    Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

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

    Science.gov (United States)

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

    2017-07-01

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

  20. Acoustoelastic evaluation of welding and heat treatment stress relieving of pressure vessel steel for Angra 3

    Energy Technology Data Exchange (ETDEWEB)

    Moraes, Bruno C. de, E-mail: bruno.cesar@nuclep.gov.br [Nuclebras Equipamentos Pesados S.A (NUCLEP), Itaguai, RJ (Brazil); Bittencourt, Marcelo de S.Q., E-mail: bruno.cesar@nuclep.gov.br, E-mail: bittenc@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Currently the knowledge of non-destructive techniques allows to evaluate the stresses on components and mechanical structures, aiming at physical security, preservation of the environment and avoid financial losses associated with the construction and operation of industrial plants. The search for new techniques, especially applied in the nuclear industry to assess status more accurately, voltage safety and to ensure structural integrity, for example, core components of the primary circuit, such as the reactor pressure vessel and steam generator has become of great importance within the community of non-destructive testing .This paper aims to contribute to the non-destructive technique development in order to ensure the structural integrity of nuclear components. One acoustoelastic evaluation of steel 20 MnMoNi 55, used in pressure vessels of nuclear power plants were performed. The acoustic birefringence technique was use to evaluate the acoustoelastic behavior of the test material in the as received condition, after welding and after the stress relief heat treatment. The constant acoustoelastic material was obtained by an uniaxial loading test. It was found a slight anisotropy in the material as received. After welding, a marked variation of acoustic birefringence in the region near the weld bead was observed. The heat treatment indicated a new change of acoustic birefringence. Obtaining the acoustoelastic constant allowed the evaluation of stress in the different conditions of the weld and treated material. (author)

  1. Acoustoelastic evaluation of welding and heat treatment stress relieving of pressure vessel steel for Angra 3

    International Nuclear Information System (INIS)

    Moraes, Bruno C. de; Bittencourt, Marcelo de S.Q.

    2015-01-01

    Currently the knowledge of non-destructive techniques allows to evaluate the stresses on components and mechanical structures, aiming at physical security, preservation of the environment and avoid financial losses associated with the construction and operation of industrial plants. The search for new techniques, especially applied in the nuclear industry to assess status more accurately, voltage safety and to ensure structural integrity, for example, core components of the primary circuit, such as the reactor pressure vessel and steam generator has become of great importance within the community of non-destructive testing .This paper aims to contribute to the non-destructive technique development in order to ensure the structural integrity of nuclear components. One acoustoelastic evaluation of steel 20 MnMoNi 55, used in pressure vessels of nuclear power plants were performed. The acoustic birefringence technique was use to evaluate the acoustoelastic behavior of the test material in the as received condition, after welding and after the stress relief heat treatment. The constant acoustoelastic material was obtained by an uniaxial loading test. It was found a slight anisotropy in the material as received. After welding, a marked variation of acoustic birefringence in the region near the weld bead was observed. The heat treatment indicated a new change of acoustic birefringence. Obtaining the acoustoelastic constant allowed the evaluation of stress in the different conditions of the weld and treated material. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Science.gov (United States)

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

    2014-02-01

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

  4. The non-destructive examination of reactor pressure vessel steels by positron annihilation

    International Nuclear Information System (INIS)

    Highton, J.P.

    1983-01-01

    The rapid radiation hardening of copper bearing reactor pressure vessel steels has been linked with microvoids that are associated with copper based complexes in the metal lattice. These microvoids are active in the sense that their size appears to be related to the temperature of irradiation, which thus determines their influence on dislocation mobility. These sites appear to grow by vacancy condensation which causes a reduction in the local lattice energy. Thus prolonged exposure to PWR temperatures, even in the absence of a neutron flux, may also cause embrittlement. It has been found that these sites, which represent a local negative charge, act as traps to positrons. The size of each site dictates its positron trapping potential. As the trapping potential increases so too does the probability that the positrons will annihilate with low momentum conduction electrons. The momentum of the annihilating electrons will determine the degree of Doppler broadening of the 511 keV annihilation gamma peak. Thus careful analysis of this peak can yield useful information on the degree of embrittlement caused by these active defect complexes. In this way positron annihilation offers a powerful non-destructive alternative to current methods of assessing the integrity of nuclear reactor pressure vessels. (author)

  5. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

  6. The influence of fire exposure on austenitic stainless steel for pressure vessel fitness-for-service assessment: Experimental research

    Science.gov (United States)

    Li, Bo; Shu, Wenhua; Zuo, Yantian

    2017-04-01

    The austenitic stainless steels are widely applied to pressure vessel manufacturing. The fire accident risk exists in almost all the industrial chemical plants. It is necessary to make safety evaluation on the chemical equipment including pressure vessels after fire. Therefore, the present research was conducted on the influences of fire exposure testing under different thermal conditions on the mechanical performance evolution of S30408 austenitic stainless steel for pressure vessel equipment. The metallurgical analysis described typical appearances in micro-structure observed in the material suffered by fire exposure. Moreover, the quantitative degradation of mechanical properties was investigated. The material thermal degradation mechanism and fitness-for-service assessment process of fire damage were further discussed.

  7. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  8. Stereofractographic investigation of static start and dynamic jump of a fatigue crack in pressure vessel steel

    International Nuclear Information System (INIS)

    Stepanenko, V.A.; Shtukaturova, A.S.; Yasnij, P.V.; AN Ukrainskoj SSR, Kiev. Inst. Problem Prochnosti)

    1983-01-01

    Results of investigaytion have been discussed ipto the effect of certain temperature-force factors on regularities in the formation of stretch zones during the crack initiation static and transition zones at crack jumps in the process of cyclic loading. The 15Kh2NMFA pressure vessel steel has been investigated. The steel fracture toughnesKub(Ic) has been determined testing s the specimens for excentric stretching or a bending through an angle. It has been shown that transition zones in a front of fatique cracks at the jump beginning and end are formed through the shift mechanism owing to the material separation along the maximum failure zone contour, i.e. along the plastic zone contour in a crack vertex. This is the mait difference of regularities in the formation of the transition zones during fatique crack jumps from stretching zones formed through the break-away mechanism of crack vertex bluntness during its static move. It is noted that a final conclusion on the mechanism of transition zone formation during fartique crack jumps allows one to perform systematic investigation into the plastic zone configuration in a fatique crack verteX and stereofractographic measurement of two identically conjugate jump surfaces on opposite fractures of the same samples

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

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

    1997-01-01

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

  11. Microstructure and mechanical characteristics of a laser welded joint in SA508 nuclear pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wei, E-mail: wei.guo-2@manchester.ac.uk [Laser Processing Research Centre, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Sackville Street, Manchester, M13 9 PL (United Kingdom); Dong, Shiyun [Laser Processing Research Centre, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Sackville Street, Manchester, M13 9 PL (United Kingdom); Institute of Laser Engineering, Beijing University of Technology, Beijing 100124 (China); Guo, Wei; Francis, John A.; Li, Lin [Laser Processing Research Centre, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Sackville Street, Manchester, M13 9 PL (United Kingdom)

    2015-02-11

    SA508 steels are typically used in civil nuclear reactors for critical components such as the reactor pressure vessel. Nuclear components are commonly joined using arc welding processes, but with design lives for prospective new build projects exceeding 60 years, new welding technologies are being sought. In this exploratory study, for the first time, autogenous laser welding was carried out on 6 mm thick SA508 Cl.3 steel sheets using a 16 kW fiber laser system operating at a power of 4 kW. The microstructure and mechanical properties (including microhardness, tensile strength, elongation, and Charpy impact toughness) were characterized and the microstructures were compared with those produced through arc welding. A three-dimensional transient model based on a moving volumetric heat source model was also developed to simulate the laser welding thermal cycles in order to estimate the cooling rates included by the process. Preliminary results suggest that the laser welding process can produce welds that are free of macroscopic defects, while the strength and toughness of the laser welded joint in this study matched the values that were obtained for the parent material in the as-welded condition.

  12. An internal-friction study of reactor-pressure-vessel steel embrittlement

    International Nuclear Information System (INIS)

    Ouytsel, K. van; Fabry, A.; Batist, R. de; Schaller, R.

    1997-01-01

    Within an enhanced commercial surveillance strategy, the nuclear-research institute SCK.CEN in Mol, Belgium is investigating, by means of internal friction, the microstructural processes responsible for embrittlement of pressure-vessel steels. The experiments were carried out using a torsion pendulum at the Ecole Polytechnique Federale de Lausanne in Switzerland. Amplitude-independent internal-friction experiments teach us that neutron irradiation induces defects which interact with mobile dislocations. Thermal ageing of JRQ and Doel-IV steel does not cause major embrittlement effects. Amplitude-dependent internal-friction experiments allow us to determine a critical amplitude which corresponds to the yield stress of the material as obtained from static tensile tests. The results also correspond to a three-component model for the yield strength taking into account both hardening and non-hardening embrittlement. Investigations of Doel-I-II weld material in different conditions reveal that embrittlement due to irradiation or thermal ageing can be interpreted in terms of a fine interplay between long- and short-range phenomena. (author)

  13. Fracture toughness behavior and its analysis on nuclear pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Iwadate, Tadao; Tanaka, Yasuhiko; Ono, Shin-ichi; Tsukada, Hisashi [Japan Steel Works Ltd., Muroran, Hokkaido. Muroran Plant

    1983-02-01

    A drop weight J sub(Id) testing machine has been developed successfully, by which the multiple specimen J resistance curve test technique can be applied to measure the fracture toughness. In this study, the use of a small size round compact tension (RCT) specimen for measuring the fracture toughness J sub(Ic) or J sub(Id) of the nuclear pressure vessel steels is recommended and confirmed for the surveillance tests. The static and dynamic fracture toughness of ASTM A508 C 1.2, A508 C 1.3 and A533 Gr.B C 1.1 steels in the wide range of temperature including the upper shelf have been measured and their behavior has been analysed. The fracture toughness behavior under various strain rates and in a wide temperature range can be explained by the behavior of stretched zone formation preceding the crack initiation. The scatter of K sub(J) values in the transition range is caused by the amount of crack extension contained in the specimens. In this paper, the method to obtain the fracture toughness equivalent to the K sub(Ic) from the K sub(J) value is also presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

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

  15. Biaxial loading effects on fracture toughness of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    McAfee, W.J.; Bass, B.R.; Bryson, J.W. Jr.; Pennell, W.E.

    1995-03-01

    The preliminary phases of a program to develop and evaluate fracture methodologies for assessing crack-tip constraint effects on fracture toughness of reactor pressure vessel (RPV) steels have been completed by the Heavy-Section Steel Technology (HSST) Program. Objectives were to investigate effect of biaxial loading on fracture toughness, quantify this effect through existing stress-based, dual-parameter, fracture-toughness correlations, or propose and verify alternate correlations. A cruciform beam specimen with 2-D, shallow, through-thickness flaw and a special loading fixture was designed and fabricated. Tests were performed using biaxial loading ratios of 0:1 (uniaxial), 0.6:1, and 1:1 (equi-biaxial). Critical fracture-toughness values were calculated for each test. Biaxial loading of 0.6:1 resulted in a reduction in the lower bound fracture toughness of ∼12% as compared to that from the uniaxial tests. The biaxial loading of 1:1 yielded two subsets of toughness values; one agreed well with the uniaxial data, while one was reduced by ∼43% when compared to the uniaxial data. Results were evaluated using J-Q theory and Dodds-Anderson (D-A) micromechanical scaling model. The D-A model predicted no biaxial effect, while the J-Q method gave inconclusive results. When applied to the 1:1 biaxial data, these constraint methodologies failed to predict the observed reduction in fracture toughness obtained in one experiment. A strain-based constraint methodology that considers the relationship between applied biaxial load, the plastic zone width in the crack plane, and fracture toughness was formulated and applied successfully to the data. Evaluation of this dual-parameter strain-based model led to the conclusion that it has the capability of representing fracture behavior of RPV steels in the transition region, including the effects of out-of-plane loading on fracture toughness. This report is designated as HSST Report No. 150

  16. Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

  17. Irradiation induced tensile property change of SA 508 Cl. 3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Kuk, Il Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were irradiated to a neutron fluence of 2.7 x 10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Ban-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation (VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by conventional TEM. (author)

  18. On the transition of short cracks into long fatigue cracks in reactor pressure vessel steels

    Directory of Open Access Journals (Sweden)

    Singh Rajwinder

    2018-01-01

    Full Text Available Short fatigue cracks, having dimension less than 1 mm, propagate at much faster rates (da/dN even at lower stress intensity factor range (da/dN as compared to the threshold stress intensity factor range obtained from long fatigue crack growth studies. These short cracks originate at the sub-grain level and some of them ultimately transit into critical long cracks over time. Therefore, designing the components subjected to fatigue loading merely on the long crack growth data and neglecting the short crack growth behavior can overestimate the component’s life. This aspect of short fatigue cracks become even more critical for materials used for safety critical applications such as reactor pressure vessel (RPV steel in nuclear plants. In this work, the transition behaviour of short fatigue crack gowth into long fatigue crack is studied in SA508 Grade 3 Class I low alloy steel used in RPVs. In-situ characterization of initiation, propagation and transition of short fatigue cracks is performed using fatigue stage for Scanning Electron Microscope (SEM in addition to digital microscopes fitted over a servo-hydraulic fatigue machine and correlated with the microtructural information obtained using electron backscatter diffraction (EBSD. SA508 steel having an upper bainitic microstructure have several microstructural interfaces such as phase and grain boundaries that play a significant role in controlling the short fatigue crack propagation. Specially designed and prepared short fatigue specimens (eletro-polished with varying initial crack lengths of the order of tens of microns are used in this study. The transition of such short initial cracks into long cracks is then tracked to give detailed insight into the role of each phase and phase/grain boundary with an objective of establishing Kitagawa-Takahashi diagram for the given RPV steel. The behavior of the transited long cracks is then compared with the crack propagation behavior obtained using

  19. Fracture toughness determination of the pressure vessel steel A508 Cl 2 between 100 and 350 degree C

    International Nuclear Information System (INIS)

    Rao, S.

    1980-09-01

    The fracture toughness of the pressure vessel steel A508 was determined in the temperature range 100 - 350 degree C. The J-integral method with crack growth resistance curves, the so-called R-curves, was used. The results show that the steel does not have an 'upper-shelf' and the fracture toughness, K sub (JC), decreases with increasing temperature to a minimum around 300 degree C and an increase above it. These results are compared to those obtained previously on an other pressure vessel steel A533B which has essentially the same temperature dependence. The results were also analysed using the Tearing modulus, T. The conclusion iw that the crack growth resistance and the crack initiation resistance (K sub (JC)) show a significant decrease around the operating temperatures as compared to 100 degree C. (author)

  20. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  1. Effects of thermal ageing on toughness properties of pressure vessel steel

    International Nuclear Information System (INIS)

    Todeschini, P.; Churier-Bossennec, H.; Massoud, J.P.; Frund, J.M.

    2015-01-01

    The reactor pressure vessel of pressurized water reactors operates at temperatures up to 325 C. degrees. The compositions and microstructures of its constitutive steel are optimized to obtain good initial toughness values and to minimize the effects of thermal ageing during service life. Intergranular segregation of embrittling elements like phosphorus is the main thermal ageing mechanism which might affect the long term toughness properties of low copper steels, despite the low diffusivity of phosphorus at the temperatures of interest. For long term operation, these effects are taken into account by prediction formulae which have been developed in the eighties and are included in the RCC-M and RSE-M codes. The presented study aims at validating these prediction formulae by exposures at moderately increased temperatures, up to 350 C. degrees, relatively to service conditions. The investigated materials are representative forgings and their welds, taking into account envelope phosphorus concentrations relatively to the French fleet. Predicted and measured embrittlement for base and weld metals are low and consistent together for the lowest phosphorus levels. The predicted effect of phosphorus content seems to be overestimated. The single coarse grain structure has been studied on one forging and shows a susceptibility to ageing similar to the fine grain one. The various heat affected zone microstructures studied with the plate having a phosphorus content of 0.017 % (fusion line, fine grains, inter-critical coarse grains) have given quite contrasted results. Inter-critical coarse grains notch positions show the lowest shifts. Code predictions are bounding the results of all considered heat affected zone microstructures with substantial margin. The increased susceptibility of heat affected zone compared to base metal seems globally overestimated

  2. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  3. Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

    International Nuclear Information System (INIS)

    Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

    1983-01-01

    This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments

  4. Heavy section steel technology program technical report No. 38. Fracture toughness characterization of HSST intermediate pressure vessel material

    International Nuclear Information System (INIS)

    Mager, T.R.; Yanichko, S.E.; Singer, L.R.

    1974-12-01

    The primary objective of the Heavy Section Steel Technology (HSST) Program is to develop pertinent fracture technology to demonstrate the structural reliability of present and contemplated water-cooled nuclear reactor pressure vessels. In order to demonstrate the ability to predict failure of large, heavy-walled pressure vessels under service type loading conditions, the fracture toughness properties of the vessel's materials must be characterized. The sampling procedure and test results are presented for vessel material supplied by the Oak Ridge National Laboratory that were used to characterize the fracture toughness of the HSST Intermediate Test Vessels. The metallurgical condition and heat treatment of the test material was representative of the vessel simulated service test condition. Test specimen locations and orientations were selected by the Oak Ridge National Laboratory and are representative of flaw orientations incorporated in the test vessels. The fracture toughness is documented for the materials from each of the eight HSST Intermediate Pressure Vessels tested to date. 7 references. (U.S.)

  5. The prediction of failure of welded steel plates for pressure vessels

    International Nuclear Information System (INIS)

    Gulvin, T.F.; Terry, P.; Webster, S.E.

    1980-01-01

    The avoidance of brittle fracture in pressure vessels and other welded steel fabrications is a matter of considerable importance. The application of fracture mechanics to this problem has led to the evolution of a philosophy which, among other things, permits estimation of tolerable crack sizes so that practical working limits can be set on inspection procedures and on material and welded joint toughness levels. The use of small-scale fracture mechanics tests, particularly crack opening displacement (COD) tests, to make the necessary material and/or joint property assessments is now commonplace. A number of possible analytical techniques have been considered. Initially, the COD data have been used with the Burdekin-Dawes design curve approach taking into account all the possible stress conditions and, it can be demonstrated, that consistently safe predictions of allowable flaw sizes can be made. Also considered, is the fracture analysis diagram approach of Harrison, Loosemore, and Milne in which a combination of fracture toughness data and mechanical property data is used to assess the probability of failure. Similarly the approach defined by Irvine and Quirk which makes use of mechanical property data without resorting to fracture mechanics and finally, a simple approach using uniaxial tensile property data only has also been examined. Comparisons have been made between these four approaches using, initially, the body of data referring to fracture tests on a single material with various defect sizes and aspect ratios, etc. and latterly, to the specific cases in the body of data on high strength steels. The results are discussed. (author)

  6. Critical experiments, measurements, and analyses to establish a crack arrest methodology for nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Hahn, G.T.

    1977-01-01

    Substantial progress was made in three important areas: crack propagation and arrest theory, two-dimensional dynamic crack propagation analyses, and a laboratory test method for the material property data base. The major findings were as follows: Measurements of run-arrest events lent support to the dynamic, energy conservation theory of crack arrest. A two-dimensional, dynamic, finite-difference analysis, including inertia forces and thermal gradients, was developed. The analysis was successfully applied to run-arrest events in DCB (double-cantilever-beam) and SEN (single-edge notched) test pieces. A simplified procedure for measuring K/sub D/ and K/sub Im/ values with ordinary and duplex DCB specimens was demonstrated. The procedure employs a dynamic analysis of the crack length at arrest and requires no special instrumentation. The new method was applied to ''duplex'' specimens to measure the large K/sub D/ values displayed by A533B steel above the nil-ductility temperature. K/sub D/ crack velocity curves and K/sub Im/ values of two heats of A533B steel and the corresponding values for the plane strain fracture toughness associated with static initiation (K/sub Ic/), dynamic initiation (K/sub Id/), and the static stress intensity at crack arrest (K/sub Ia/) were measured. Possible relations among these toughness indices are identified. During the past year the principal investigators of the participating groups reached agreement on a crack arrest theory appropriate for the pressure vessel problem. 7 figures

  7. Empirical correlation between mechanical and physical parameters of irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Tipping, P.; Solt, G.; Waeber, W.

    1991-02-01

    Neutron irradiation embrittlement of nuclear reactor pressure vessel (PV) steels is one of the best known ageing factors of nuclear power plants. If the safety limits set by the regulators for the PV steel are not satisfied any more, and other measures are too expensive for the economics of the plant, this embrittlement could lead to the closure of the plant. Despite this, the fundamental mechanisms of neutron embrittlement are not yet fully understood, and usually only empirical mathematical models exist to asses neutron fluence effects on embrittlement, as given by the Charpy test for example. In this report, results of a systematic study of a French forging (1.2 MD 07 B), irradiated to several fluences will be reported. Mechanical property measurements (Charpy tensile and Vickers microhardness), and physical property measurements (small angle neutron scattering - SANS), have been done on specimens having the same irradiation or irradiation-annealing-reirradiation treatment histories. Empirical correlations have been established between the temperature shift and the decrease in the upper shelf energy as measured on Charpy specimens and tensile stresses and hardness increases on the one hand, and the size of the copper-rich precipitates formed by the irradiation on the other hand. The effect of copper (as an impurity element) in enhancing the degradation of mechanical properties has been demonstrated; the SANS measurements have shown that the size and amount of precipitates are important. The correlations represent the first step in an effort to develop a description of neutron irradiation induced embrittlement which is based on physical models. (author) 6 figs., 27 refs

  8. On the composition and structure of nanoprecipitates in irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.; Liu, C.L.; Wirth, B.D.

    1997-01-01

    Nanoscale Cu rich precipitates (CRPs) are widely believed to be the dominant hardening feature resulting in severe embrittlement in irradiated reactor pressure vessel (RPV) steels. However, this view has recently been challenged by interpretations of atom probe field ion microscopy (APFIM) measurements that describe the dominant nanofeatures as dilute solute atmospheres (DSAs). The practical impact of these differing views is very significant. This work compares and contrasts the CRP versus DSA descriptions to a wide variety of pertinent data. Mechanical property trends as well as small angle neutron scattering (SANS) and field emission scanning transmission electron microscopy (FEGSTEM) measurements support the presence of CRPs. CRPs are also consistent with the fundamental thermodynamic and kinetic laws. However, standard theory cannot provide the atomic level resolution needed to fully understand the nanofeatures. Therefore, a new Lattice Monte Carlo (LMC) atomistic method is used to simulate the complex chemical structures of the CRPs. The LMC method unifies the SANS/FEGSTEM and APFIM data within a well founded physical framework

  9. The response of pressure vessel steel specimens on drop weight loading

    International Nuclear Information System (INIS)

    Winkler, S.; Kalthoff, J.F.; Gerscha, A.

    1979-01-01

    Load records obtained in instrumented impact tests in general are disturbed by inertia effects. The influence of mechanical damping provisions on these disturbing inertia effects is investigated. Precracked bend specimens are dynamically loaded in a drop weight testing system. The specimens of size 620 mm x 150 mm (25 mm or 50 mm thick) were machined from the pressure vessel steel 22 NiMoCr 37 which was heat treated to achieve a specially hardened condition. The tests were performed at two different low temperatures. The impact velocity was about 4 m/s. As it is usual in instrumented impact testing, the load at the tup of the impining striker is recorded as a function of time during the impact process. In addition the specimen is instrumented by a strain gage close to the crack tip in order to directly measure the stress intensification. Experiments were performed under pure and damped impact conditions. Damping was achieved by utilizing a soft aluminum plate between the striker and the specimen. (orig.)

  10. Postirradiation recovery of a reactor pressure vessel steel investigated by positron annihilation and microhardness measurements

    International Nuclear Information System (INIS)

    Pareja, R.; Diego, N. De; Cruz, R.M. de la; Del Rio, J.

    1993-01-01

    Positron lifetime and microhardness measurements have been performed on untreated, thermal-aged, neutron-irradiated, and postirradiation-annealed samples of reactor pressure vessel steels with the purpose of investigating the mechanisms of irradiation-induced hardening and recovery of the mechanical properties in these materials. The positron lifetime experiments have not revealed any evidence of the formation of a significant concentration of voids or vacancy clusters in samples irradiated at ∼290 C with fluences ≤2.71 x 10 23 n/m 2 (E>1 MeV), but they suggest a dislocation annealing induced by the irradiation. Isochronal annealing experiments with neutron-irradiated samples show a simultaneous recovery in their positron lifetime and microhardness at ∼340 C. From the microhardness measurements, the yield strength of the irradiated material has been estimated. The results appear to be consistent with a model of hardening due to irradiation-induced dissolution of precipitates with formation of small metastable precipitates after postirradiation aging and recovery induced by the disappearance of these metastable precipitates

  11. Appropriate welding conditions of temper bead weld repair for SQV2A pressure vessel steel

    International Nuclear Information System (INIS)

    Mizuno, R.; Matsuda, F.; Brziak, P.; Lomozik, M.

    2004-01-01

    Temper bead welding technique is one of the most important repair welding methods for large structures for which it is difficult to perform the specified post weld heat treatment. In this study, appropriate temper bead welding conditions to improve the characteristics of heat affected zone (HAZ) are studied using pressure vessel steel SQV2A corresponding to ASTM A533 Type B Class 1. Thermal/mechanical simulator is employed to give specimens welding thermal cycles from single to quadruple cycle. Charpy absorbed energy and hardness of simulated CGHAZ by first cycle were degraded as compared with base metal. Improvability of these degradations by subsequent cycles is discussed and appropriate temper bead thermal cycles are clarified. When the peak temperature lower than Ac1 and near Ac1 in the second thermal cycle is applied to CGAHZ by first thermal cycle, the characteristics of CGHAZ improve enough. When the other peak temperatures (that is, higher than Ac1) in the second thermal cycle are applied to the CGHAZ, third or more thermal cycle temper bead process should be applied to improve the properties. Appropriate weld condition ranges are selected based on the above results. The validity of the selected ranges is verified by the temper bead welding test. (orig.)

  12. The elevated temperature and thermal shock fracture toughnesses of nuclear pressure vessel steel

    International Nuclear Information System (INIS)

    Hirano, Kazumi; Kobayashi, Hideo; Nakazawa, Hajime; Nara, Atsushi.

    1979-01-01

    Thermal shock experiments were conducted on nuclear pressure vessel steel A533 Grade B Class 1. Elastic-plastic fracture toughness tests were carried out within the same high temperature range of the thermal shock experiment and the relation between stretched zone width, SZW and J-integral was clarified. An elastic-plastic thermal shock fracture toughness value. J sub(tsc) was evaluated from a critical value of stretched zone width, SZW sub(tsc) at the initiation of thermal shock fracture by using the relation between SZW and J. The J sub(tsc) value was compared with elastic-plastic fracture toughness values, J sub( ic), and the difference between the J sub(tsc) and J sub( ic) values was discussed. The results obtained are summarized as follows; (1) The relation between SZW and J before the initiation of stable crack growth in fracture toughness test at a high temperature can be expressed by the following equation regardless of test temperature, SZW = 95(J/E), where E is Young's modulus. (2) Elevated temperature fracture toughness values ranging from room temperature to 400 0 C are nearly constant regardless of test temperature. It is confirmed that upper shelf fracture toughness exists. (3) Thermal shock fracture toughness is smaller than elevated temperature fracture toughness within the same high temperature range of thermal shock experiment. (author)

  13. Heavy-Section Steel Technology Program intermediate-scale pressure vessel tests

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Smith, G.C.; Whitman, G.D.

    1977-01-01

    The tests of intermediate-size vessels with sharp flaws permitted the comparison of experimentally observed behavior with analytical predictions of the behavior of flawed pressure vessels. Fracture strains estimated by linear elastic fracture mechanics (LEFM) were accurate in the cases in which the flaws resided in regions of high transverse restraint and the fracture toughness was sufficiently low for unstable fracture to occur prior to yielding through the vessel wall. When both of these conditions were not present, unstable fracture did occur, always preceded by stable crack growth; and the cylinders with flaws initially less than halfway through the wall attained gross yield prior to burst. Predictions of failure pressure of the vessels with flawed nozzles, based upon LEFM estimates of failure strain, were very conservative. LEFM calculations of critical load were based upon small-specimen fracture toughness test data. Whenever gross yielding preceded failure, the actual strains achieved were considerably greater than the estimated strains at failure based on LEFM. In such cases the strength of the vessel may be no longer dependent upon plane-strain fracture toughness but upon the capacity of the cracked section to carry the imposed load stably in the plastic range. Stable crack growth, which has not been predictable quantitatively, is an important factor in elastic-plastic analysis of strength. The ability of the flawed vessels to attain gross yield in unflawed sections has important qualitative implications on pressure vessel safety margins. The gross yield condition occurs in light-water-reactor pressure vessels at about 2 x design pressure. The intermediate vessel tests that demonstrated a capacity for exceeding this load confirm that the presumed margin of safety is not diminished by the presence of flaws of substantial size, provided that material properties are adequate

  14. Potential high fluence response of pressure vessel internals constructed from austenitic stainless steels

    International Nuclear Information System (INIS)

    Garner, F.A.; Greenwood, L.R.; Harrod, D.L.

    1993-08-01

    Many of the in-core components in pressurized water reactors are constructed of austenitic stainless steels. The potential behavior of these components can be predicted using data on similar steels irradiated at much higher displacement rates in liquid-metal reactors or water-cooled mixed-spectrum reactors. Consideration of the differences between the pressurized water environment and that of the other reactors leads to the conclusion that significant amounts of void swelling, irradiation creep, and embrittlement will occur in some components, and that the level of damage per atomic displacement may be larger in the pressurized water environment

  15. Applicability of the fracture toughness master curve to irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Sokolov, M.A.; McCabe, D.E.; Alexander, D.J.; Nanstad, R.K.

    1997-01-01

    The current methodology for determination of fracture toughness of irradiated reactor pressure vessel (RPV) steels is based on the upward temperature shift of the American Society of Mechanical Engineers (ASME) K Ic curve from either measurement of Charpy impact surveillance specimens or predictive calculations based on a database of Charpy impact tests from RPV surveillance programs. Currently, the provisions for determination of the upward temperature shift of the curve due to irradiation are based on the Charpy V-notch (CVN) 41-J shift, and the shape of the fracture toughness curve is assumed to not change as a consequence or irradiation. The ASME curve is a function of test temperature (T) normalized to a reference nit-ductility temperature, RT NDT , namely, T-RT NDT . That curve was constructed as the lower boundary to the available K Ic database and, therefore, does not consider probability matters. Moreover, to achieve valid fracture toughness data in the temperature range where the rate of fracture toughness increase with temperature is rapidly increasing, very large test specimens were needed to maintain plain-strain, linear-elastic conditions. Such large specimens are impractical for fracture toughness testing of each RPV steel, but the evolution of elastic-plastic fracture mechanics has led to the use of relatively small test specimens to achieve acceptable cleavage fracture toughness measurements, K Jc , in the transition temperature range. Accompanying this evolution is the employment of the Weibull distribution function to model the scatter of fracture toughness values in the transition range. Thus, a probabilistic-based bound for a given data population can be made. Further, it has been demonstrated by Wallin that the probabilistic-based estimates of median fracture toughness of ferritic steels tend to form transition curves of the same shape, the so-called ''master curve'', normalized to one common specimen size, namely the 1T [i.e., 1.0-in

  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. Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

    2014-01-01

    The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

  18. Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

    International Nuclear Information System (INIS)

    Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

    1985-01-01

    This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed at ORNL for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along wih applications to pressure vessel experiments. (orig./HP)

  19. Fatigue crack growth behavior of pressure vessel steels and submerged arc weldments in a high-temperature pressurized water environment

    International Nuclear Information System (INIS)

    Liaw, P.K.; Logsdon, W.A.; Begley, J.A.

    1989-01-01

    The fatigue crack growth rate (FCGR) properties of SA508 Cl 2a and SA533 Gr A Cl 2 pressure vessel steels and the corresponding automatic submerged arc weldments were developed in a high-temperature pressurized water (HPW) environment at 288 degrees C (550 degrees F) and 7.2 MPa (1044 psi) at load ratios of 0.20 and 0.50. The properties were generally conservative compared to American Society of Mechanical Engineers Section XI water environment reference curve. The growth rate of fatigue cracks in the base materials, however, was faster in the HPW environment than in a 288 degrees C (550 degrees F) base line air environment. The growth rate of fatigue cracks in the two submerged arc weldments was also accelerated in the HPW environment but to a lesser degree than that demonstrated by the base materials. In the air environment, fatigue striations were observed, independent of material and load ratio, while in the HPW environment, some intergranular facets were present. The greater environmental effect on crack growth rates displayed by the base materials compared the weldments attributed to a different sulfide composition and morphology

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

    Energy Technology Data Exchange (ETDEWEB)

    Li, C.W.; Han, L.Z.; Luo, X.M.; Liu, Q.D.; Gu, J.F., E-mail: gujf@sjtu.edu.cn

    2016-08-15

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe{sub 3}C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo{sub 2}C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained. - Highlights: • The dependence of the deterioration of impact toughness on tempering temperature has been analysed. • The instrumented Charpy V-notch impact test has been employed to study the fracture mechanism. • The influence of M/A constituents on different fracture mechanisms based on the hinge model has been demonstrated. • A correlation between the mechanical properties and the amount of M/A constituents has been established.

  1. Proof, interpretation and evaluation of radiation-induced microstructural changes in WWER reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Boehmert, J.; Gokhman, A.; Grosse, M.; Ulbricht, A.

    2003-06-01

    Neutron embrittlement is a special issue for the VVER-type reactors. One of the fundamentals for a reliable assessment of the current material state is knowledge of the causes and mechanisms of neutron embrittlement. The aim of the project is to understand and to quantify the microstructural appearances due to neutron radiation in VVER-type reactor pressure vessel steels. The material base is a broad variation of irradiation probes taken from the irradiation programme Rheinsberg, surveillance programmes of Russian, Ukrainian or Hungarian NPPs or irradiation experiments with mockup-alloys. The microstructure was investigated by different methods. The small angle neutron scattering (SANS) proved to be the most suitable method. A procedure was developed to determine mean diameter, size distribution and volume fraction of irradiation-induced microstructure from SANS experiments in a reliable and comparable manner. With this method microstructural parameters were systematically determined and the main factors of influence were identified. Apart from the neutron fluence the volume fraction of radiation defects mainly changes with the copper or nickel content whereas phosphorus is hardly relevant. Annealing remedies the radiation-induced microstructural appearances. The ratio between nuclear and magnetic neutron scattering provides information on the type of radiation defects. This leads to the conclusion that the material composition changes the radiation defects. The change occurs gradually rather than abruptly. The radiation defects detected by SANS correlate with the radiation hardening and embrittlement. Generally, the results suggest a bimodal mechanism due to radiation-enhanced and radiation-induced defect evolution. A kinetic model on base of the rate theory approach was established. (orig.)

  2. Long-term irradiation effects on reactor-pressure vessel steels. Investigations on the nanometer scale

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, Arne

    2017-06-01

    The exposure of reactor pressure vessel (RPV) steels to neutron irradiation gives rise to irradiation-enhanced diffusion, a rearrangement of solute atoms and, consequently, a degradation of the mechanical properties. The increasing age of existing nuclear power plants raises new questions specific to long-term operation. Two of them are addressed in this thesis: flux effects and the late-blooming effect. Can low-flux irradiations up to a given fluence be reproduced by more rapid high-flux irradiations up to the same fluence? Can the irradiation response of RPV steels be extrapolated to higher fluences or are there unexpected ''late-blooming'' effects. Small-angle neutron scattering (SANS), atom-probe tomography (APT) and Vickers-hardness testing were applied. A novel Monte-Carlo based fitting algorithm for SANS data was implemented in order to derive statistically reliable characteristics of irradiation-induced solute-atom clusters. APT was applied in selected cases to gain additional information on the composition and the shape of clusters. Vickers hardness testing was performed on the SANS samples to link the nanometer-scale changes to irradiation hardening. The investigations on flux effects show that clusters forming upon high-flux irradiation are smaller and tend to have a higher number density compared to low-flux irradiations at a given neutron fluence. The measured flux dependence of the cluster-size distribution is consistent with the framework of deterministic growth (but not with coarsening) in combination with radiation-enhanced diffusion. Since the two effects on cluster-size and volume fraction partly cancel each other out, no significant effect on the hardening is observed. The investigations of a possible late-blooming effect indicate that the very existence (yes or no) of such an effect depends on the irradiation conditions. Irradiations at lower fluxes and a lower temperature (255 C) give rise to a significant increase of the

  3. Studies on the welding of heavy-section ASTM A542 Cl. 1 steel for large-sized pressure vessels

    International Nuclear Information System (INIS)

    Shimizu, Shigeki; Aota, Toshiichi; Kasahara, Masayuki

    1977-01-01

    ASTM A 542, Cl. 1 steel was developed and standardized recently, and is excellent in the high temperature strength and toughness as compared with conventionally used A 387, Grade 22 steel, accordingly the application to large pressure vessels is planned. This steel is a low alloy steel, and in case of large thickness, the possibility of cracking in the welded part is large. Also many times of annealing are required for the prevention of welding cracking, the relieving of residual stress, and the softening of hardened portion, but the possibility of cracking during stress-relieving annealing is large. In this study, Tekken type cracking test was carried out by coated electrode welding, and restricted cracking test was carried out by submerged arc welding of the A 542, Cl. 1 steel and A 387, Grade 22 steel, thus the welding cracking property was investigated, and the optimal welding conditions were selected. Also the test of cracking during the stress-relieving annealing of both steels was carried out, and the method of preventing the cracking was studied. The optimal conditions of stress-relieving annealing were selected, and the mechanism of the cracking was clarified. The mechanical properties of the joints welded and stress-relieved under the selected conditions were confirmed. (Kako, I.)

  4. Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

    International Nuclear Information System (INIS)

    Blagoeva, D.T.; Debarberis, L.; Jong, M.; Pierick, P. ten

    2014-01-01

    This paper is illustrating the potential of the well-known low alloyed clean steels, extensively used for the current light water Reactor Pressure Vessels (RPV) steels, for a likely use as a structural material also for the new generation nuclear systems. This option would provide, especially for large components, affordable, easily accessible and a technically more convenient solution in terms of manufacturing and joining techniques. A comprehensive comparison between several sets of surveillance and research data available for a number of RPV clean steels for doses up to 1.5 dpa, and up to 12 dpa for 9%Cr steels, is carried out in order to evaluate radiation stability of the currently used RPV clean steels even at higher doses. Based on the numerous data available, positive preliminary conclusions are drawn regarding the eventual use of clean RPV steels for the massive structural components of the new reactor systems. - Highlights: • Common embrittlement trend between RPV and advanced steels till intermediate doses. • For doses >1.5 dpa, damage rate saturation tendency is observed for RPV steels. • RPV steels might be conveniently utilised also outside their foreseen dose range

  5. 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

  6. Application of the master curve approach to fracture mechanics characterisation of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Viehrig, Hans-Werner; Zurbuchen, Conrad; Kalkhof, Dietmar

    2010-06-01

    The paper presents results of a research project founded by the Swiss Federal Nuclear Inspectorate concerning the application of the Master Curve approach in nuclear reactor pressure vessels integrity assessment. The main focus is put on the applicability of pre-cracked 0.4T-SE(B) specimens with short cracks, the verification of transferability of MC reference temperatures T 0 from 0.4T thick specimens to larger specimens, ascertaining the influence of the specimen type and the test temperature on T 0 , investigation of the applicability of specimens with electroerosive notches for the fracture toughness testing, and the quantification of the loading rate and specimen type on T 0 . The test material is a forged ring of steel 22 NiMoCr 3-7 of the uncommissioned German pressurized water reactor Biblis C. SE(B) specimens with different overall sizes (specimen thickness B=0.4T, 0.8T, 1.6T, 3T, fatigue pre-cracked to a/W=0.5 and 20% side-grooved) have comparable T 0 . T 0 varies within the 1σ scatter band. The testing of C(T) specimens results in higher T 0 compared to SE(B) specimens. It can be stated that except for the lowest test temperature allowed by ASTM E1921-09a, the T 0 values evaluated with specimens tested at different test temperatures are consistent. The testing in the temperature range of T 0 ± 20 K is recommended because it gave the highest accuracy. Specimens with a/W=0.3 and a/W=0.5 crack length ratios yield comparable T 0 . The T 0 of EDM notched specimens lie 41 K up to 54 K below the T 0 of fatigue pre-cracked specimens. A significant influence of the loading rate on the MC T 0 was observed. The HSK AN 425 test procedure is a suitable method to evaluate dynamic MC tests. The reference temperature T 0 is eligible to define a reference temperature RT To for the ASME-KIC reference curve as recommended in the ASME Code Case N-629. An additional margin has to be defined for the specific type of transient to be considered in the RPV integrity assessment

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-26

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

  9. Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

    International Nuclear Information System (INIS)

    Ishino, Shiori

    2007-01-01

    The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

  10. Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

    1991-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

  11. Computational evaluation of the constraint loss on the fracture toughness of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Serrano Garcia, M.

    2007-01-01

    The Master Curve approach is included on the ASME Code through some Code Cases to assess the reactor pressure vessel integrity. However, the margin definition to be added is not defined as is the margin to be added when the Master Curve reference temperature T 0 is obtained by testing pre-cracked Charpy specimens. The reason is that the T 0 value obtained with this specimen geometry is less conservative than the value obtained by testing compact tension specimens possible due to a loss of constraint. The two parameter fracture mechanics, considered as an extension of the classical fracture mechanics, coupled to a micromechanical fracture models is a valuable tool to assess the effect of constraint loss on fracture toughness. The definition of a parameter able to connect the fracture toughens value to the constraint level on the crack tip will allow to quantify margin to be added to the T 0 value when this value is obtained testing the pre-cracked Charpy specimens included in the surveillance capsule of the reactor pressure vessel. The Nuclear Regulatory Commission (NRC) define on the To value obtained by testing compact tension specimens and ben specimens (as pre-cracked Charpy are) bias. the NRC do not approved any of the direct applications of the Master Curve the reactor pressure vessel integrity assessment until this bias will be quantified in a reliable way. the inclusion of the bias on the integrity assessment is done through a margin to be added. In this thesis the bias is demonstrated an quantified empirical and numerically and a generic value is suggested for reactor pressure vessel materials, so that it can be used as a margin to be added to the T 0 value obtained by testing the Charpy specimens included in the surveillance capsules. (Author) 111 ref

  12. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  13. Stress-relieving annealing of Cr-Mo steel for high temperature pressure vessels and the quality change in use

    International Nuclear Information System (INIS)

    Makioka, Minoru; Hirano, Hiromichi

    1976-01-01

    The securing of good mechanical properties is difficult in thick plates for large pressure vessels because cooling rate is insufficient and time is prolonged in heat treatment. Cr-Mo steel plates are usually used in the state of improved notch toughness though somewhat reduced strength by normalizing or accelerated cooling and tempering. If the time for heat treatment is prolonged, the embrittlement occurs. The effects of temperature, holding time, and cooling rate in stress-relieving treatment on the mechanical properties of 1-1/4Cr - 1/2Mo, 2-1/4Cr - 1Mo, 3Cr - 1Mo, and 5Cr - 1/2Mo steels were investigated. The tensile strength lowered almost linearly as the hollomon-Jaffe parameter of heat treatment condition increased in all the steels. The transition temperature shifted continuously to high temperature side in 1-1/4Cr - 1/2Mo steel, but the notch toughness was improved up to certain values and then the tendency turning to brittleness was shown in the other steels, as the H-J parameter increased. When the holding time became longer, the transition temperature shifted to higher temperature side, but the cooling rate showed no effect. The condition for stress relieving treatment must be selected so that the ferrite bands observed in welded metal do not arise. The embrittlement at the operation temperature of 400 - 450 0 C for a long time is evaluated by the comparison with that by stepped cooling method. (Kako, I.)

  14. Creep deformation and crack growth in a low alloy steel welded pressure vessel containing defects

    International Nuclear Information System (INIS)

    Coleman, M.C.

    1982-01-01

    A full-size pressure vessel was tested for effects of welding residual stresses on creep deformation and crack growth. The vessel, based on 1/2 Cr 1/2 Mo 1/4 V main steam pipe, contained four 2CrMo manual metal arc welds, two in the as-welded condition and two stress-relieved. All the welds contained pre-existing defects machined in the heat affected zones. Testing was carried out at two internal steam pressures, 250 and 350 bar, and 565 0 C. Cracked and uncracked areas of the vessel were monitored continuously. Results are presented for the continuous creep deformation observed in both the hoop and axial directions of the welds throughout the 11,400 h of testing, as well as the intermittent strain data obtained during inspections. Crack growth observations are described based on nondestructive examination. The residual stresses measured are also given for both the as-welded and stress relieved weldments. Results obtained are discussed in terms of the effects of welding residual stress on the hoop and axial deformations observed in the welds. Similarly, the effects of residual stress on creep crack growth are considered together with compositional and microstructural implications. 9 figures, 5 tables

  15. A phenomenological method of mechanical properties definition of reactor pressure vessels (RPV) steels VVER according to the ball indentation diagram

    International Nuclear Information System (INIS)

    Bakirov, M. B.; Potapov, V.V.; Massoud, J.P.

    2002-01-01

    This work presents specimen-free methods of a standard uniaxial tension diagram construction and RPV (reactor pressure vessel) steels VVER strength properties definition out of a continuous ball indentation diagram. A similarity phenomenon of uniaxial tension strain curves at a hardening area and an area of a ball indentation constitutes the ground of the methods. The methods are developed on the basis of the uniform graphic representation of elasto-plastic strain processes by indentation and tension and with the reception of the unified yield curve at a hardening area. The calculation results on the phenomenological method conducted for a wide range of RPV steels conditions of nuclear reactors have shown a good precision as far as strain curves construction by the uniaxial tension out of the elasto-plastic indentation diagram is concerned. (authors)

  16. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    International Nuclear Information System (INIS)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi

    2017-01-01

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t_8_/_5 (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  17. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi [Inner Mongolia Univ. of Science and Technology, Baotou (China). School of Material and Metallurgy; Kang, Xiaolan [Baotou Vocational and Technical College (China)

    2017-02-15

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t{sub 8/5} (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  18. Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

    Science.gov (United States)

    Krasikov, E.; Nikolaenko, V.

    2017-01-01

    Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

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

    International Nuclear Information System (INIS)

    Kleber, X.; Saillet, S.

    2011-01-01

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

  20. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  1. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

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

  2. Inelastic Cyclic Deformation Behaviors of Type 316H Stainless Steel for Reactor Pressure Vessel of Sodium-Cooled Fast Reactor at Elevated Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji-Hyun; Hong, Seokmin; Koo, Gyeong-Hoi; Lee, Bong-Sang; Kim, Young-Chun

    2015-01-01

    Type 316H stainless steel is a primary candidate material for a reactor pressure vessel of a sodium-cooled fast (SFR) reactor which is under development in Korea. The reactor pressure vessel for a SFR is subjected to inelastic deformation induced by cyclic thermal stress. Fully reversed cyclic testing and ratcheting testing at elevated temperatures were performed to characterize the inelastic cyclic deformation behaviors of Type 316H stainless steel at the SFR operating temperature. It was found that cyclic hardening of Type 316H stainless steel was enhanced, and the accumulation of ratcheting deformation of Type 316H stainless steel was retarded at around the SFR operating temperature. The results of the tensile testing and the microstructural investigation for dislocated structures after the inelastic deformation testing showed that dynamic strain aging affected the inelastic cyclic deformation behavior of Type 316 stainless steel at around the SFR operating temperature.

  3. Overview of French activities on neutron radiation embrittlement of pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Brillaud, C [Electricite de France (EDF), 37 - Tours (France); Keroulas, F de [Electricite de France (EDF), 93 - Saint-Denis (France); Pichon, C [Electricite de France (EDF), 69 - Villeurbanne (France); Teissier, A [Electricite de France (EDF), 92 - Courbevoie (France). Service Etudes et Projets Thermiques et Nucleaires

    1994-12-31

    This paper describes recent developments in France`s pressure vessel surveillance program, particularly aimed at assessing the irradiation-caused embrittlement of EDF`s PWRs. The first part presents surveillance program results for base metal, weld metal and heat-affected zones for 74 capsules removed from 34 units. Fluence ranges from 0.3.10{sup 19} n.cm{sup -2} to 5.5.10{sup 19} n.cm{sup -2}. The second part considers research and development activities in this area: these include the metallurgical structure effects of segregated bands on mechanical properties and the embrittlement rate under irradiation, as well as the effect of irradiation parameters such as flux and neutron spectrum on irradiation embrittlement, and more especially to obtain the best damage assessment. (authors). 14 refs., 5 figs., 1 tab.

  4. Formation of microcracks during stress-relief annealing of a weldment in pressure vessel steel of type A508 C1 2

    International Nuclear Information System (INIS)

    Liljestrand, L.-G.; Oestberg, G.; Lindhagen, P.

    1978-01-01

    Crack formation in the heat-affected zones of heavy section weldments of type A 508 C1 2 pressure vessel steel during stress-relief annealing has been studied on an actual weldment and on simulated structures. Mechanical testing of the latter showed that stress relaxation of the order of magnitude occuring during stress-relief annealing can produce cracks of the same kind as occasionally found in weldments of pressure vessel steel. The primary cause is believed to be grain boundary sliding, possibly but not necessarily enhanced by impurities. (Auth.)

  5. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

    Directory of Open Access Journals (Sweden)

    E. A. Kuleshova

    2017-01-01

    Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

  6. Development of a crack monitoring technique for use in a corrosion fatigue study of SA533-B pressure vessel steel

    International Nuclear Information System (INIS)

    Benson, J.M.; Tait, R.B.; Garrett, G.G.

    1981-10-01

    At present there does not exist a realistic crack growth law which will provide a good description of the relationship between the alternating stress intensity factor and the crack growth per cycle of stress. Such a law should be applicable to either the pressurized water reactor environment (PWR) or boiling water reactor environmnt (BWR). This project was formulated with the aim of examining the fatigue crack growth rate of SA533-B steel (a nuclear pressure vessel steel) in the threshold region in a simulated PWR environment. The aim of this report is to develop a crack monitoring technique for use in corrosion fatigue studies. Factors affecting fatigue crack propagation include: frequency, stress range, the effect of irradiation, ageing and environment. The mechanisms of crack propagation that are discussed include: slip dissolution, hydrogen assisted cracking, corrosion potential, and morphology studies. D.C. electrical potential, the compliance technique and the back-faced strain gauge method can be used for crack monitoring. Details are also given on the experimental equipment and programme. The results of the experiment has shown that the potential difference technique for monitoring crack length is a valuable one and is well suited for use in fatigue testing applications

  7. Comparative assessment of cyclic J-R curve determination by different methods in a pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Chowdhury, Tamshuk, E-mail: tamshuk@gmail.com [Deep Sea Technologies, National Institute of Ocean Technology, Chennai, 600100 (India); Sivaprasad, S.; Bar, H.N.; Tarafder, S. [Fatigue & Fracture Group, Materials Science and Technology Division, CSIR-National Metallurgical Laboratory, Jamshedpur, 831007 (India); Bandyopadhyay, N.R. [School of Materials Science and Engineering, Indian Institute of Engineering, Science and Technology, Shibpur, Howrah, 711103 (India)

    2016-04-15

    Cyclic J-R behaviour of a reactor pressure vessel steel using different methods available in literature has been examined to identify the best suitable method for cyclic fracture problems. Crack opening point was determined by moving average method. The η factor was experimentally determined for cyclic loading conditions and found to be similar to that of ASTM value. Analyses showed that adopting a procedure analogous to the ASTM standard for monotonic fracture is reasonable for cyclic fracture problems, and makes the comparison to monotonic fracture results straightforward. - Highlights: • Different methods of cyclic J-R evaluation compared. • A moving average method for closure point proposed. • η factor for cyclic J experimentally validated. • Method 1 is easier, provides a lower bound and direct comparison to monotonic fracture.

  8. On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Lambrecht, M.; Meslin, E.; Malerba, L.; Hernandez-Mayoral, M.; Bergner, F.; Pareige, P.; Radiguet, B.; Almazouzi, A.

    2010-01-01

    A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ∼7 x 10 19 n cm -2 . This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J. [Centre de l``Etude de l``Energie Nucleaire, Mol (Belgium); Biemiller, E.C. [Yankee Atomic Electric Company, Bolton (United States); Rossinski, S.T.; Carter, R.G. [Electric Power Research Institute, Charlotte (United States)

    1996-07-01

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

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

    International Nuclear Information System (INIS)

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

    1996-07-01

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

  11. Historical summary of the heavy-section steel technology program and some related activities in light-water reactor pressure vessel safety research

    International Nuclear Information System (INIS)

    Whitman, G.D.

    1986-03-01

    The accomplishments of the Heavy-Section Steel Technology Program and other programs having a close relationship to the development of information used in the assessment of light-water reactor pressure vessel integrity are reviewed. The early Pressure Vessel Research Committee planning, the principals contributing to program formulation, the role of the US Atomic Energy Commission, and the developments under the US Nuclear Regulatory Commission sponsorship are identified. The need for major research and development accomplishments in fracture mechanics, heavy-section steel procurement, materials properties, irradiation effects, fatigue crack growth, and structural testing are summarized. The impact of program results on regulatory issues and the development of data used in the preparation of codes, standards, and guides are discussed. Continuing activities and recommendations for future research and development in support of pressure vessel integrity assessments are presented

  12. Models for embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1995-05-01

    The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  13. In-service inspection as an aid to steel pressure vessel reliability

    International Nuclear Information System (INIS)

    Nichols, R.W.

    1975-01-01

    In-service inspection has played an important role in non-nuclear pressure vessel technology, being a legal requirement in many countries. Evidence from surveys of reliability of non-nuclear plant has suggested that such inspections can be effective in reducing the risk of subsequent failures. Recent requirements of the ASME XI code which will be summarised have important implications on the techniques to be used for in-service inspection, and so on design and fabrication aspects. Moreover, in-service inspection can only be an effective procedure if its possible weaknesses are recognised. The first problem is to ensure that an ultrasonic technique is used which is capable of detecting defects of an order of magnitude smaller than the critical size for each particular situation, in whatever defect orientation is important. The potential of different ultrasonic techniques will be compared. Next it is necessary to ensure coverage of all the relevant material. In this respect machine operation is superior to manual scanning, so that manipulation and scanning devices have to be developed. Problems of local geometry and of deviations in geometry have to be discussed with designer and fabricator; plate and clad quality have to be controlled (with respect to surface contour, metallurgical condition and freedom from interfering defects) to ensure inspectability in depth. The reliability of the mechanical and electronic equipment has to be assessed and designed to meet high requirements. Some presentational aids to detection and interpretation will be discussed. Having located a potential defect, the application of fracture mechanics treatments requires knowledge of size, shape and orientation. Some of the problems will be discussed together with possible solutions. (author)

  14. Comparison of SA508 Gr.3 and SA508 Gr.4N Low Alloy Steels for Reactor Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S

    2009-12-15

    The microstructural characteristics and mechanical properties of SA508 Gr.3 Mn-Mo-Ni low alloy steel and SA508 Gr.4N Ni-Mo-Cr low alloy steel were investigated. The differences in the stable phases between these two low alloy steels were evaluated by means of a thermodynamic calculation using ThermoCalc. They were then compared to microstructural features and correlated with mechanical properties. Mn-Mo-Ni low alloy steel shows the upper bainite structure which has the coarse cementite in the lath boundaries. However, Ni-Mo-Cr low alloy steel shows the mixture of lower bainite and tempered martensite structure that homogeneously precipitates the small carbides such as M{sub 23}C{sub 6} and M{sub 7}C{sub 3} due to an increase of hardenability and Cr addition. In the mechanical properties, Ni-Mo-Cr low alloy steel has higher strength and toughness than Mn-Mo-Ni low alloy steel. Ni and Cr additions increase the strength by solid solution hardening. Besides, microstructural changes from upper bainite to tempered martensite improve the strength of the low alloy steel by grain refining effect. And the changes in the precipitation behavior by Cr addition improve the ductile-brittle transition behavior along with a toughening effect of Ni addition.

  15. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  16. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Oh, Y. J.

    2003-01-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T sp ), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T 41J . T sp from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T 0 ) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen

  17. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  18. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  19. 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

  20. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, A.V., E-mail: Alexey.V.Subbotin@gmail.com [Scientific and Production Complex Atomtechnoprom, Moscow 119180 (Russian Federation); Panyukov, S.V., E-mail: panyukov@lpi.ru [PN Lebedev Physics Institute, Russian Academy of Sciences, Moscow 117924 (Russian Federation)

    2016-08-15

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  1. The Investigation on Strain Strengthening Induced Martensitic Phase Transformation of Austenitic Stainless Steel: A Fundamental Research for the Quality Evaluation of Strain Strengthened Pressure Vessel

    Science.gov (United States)

    Li, Bo; Cai Ren, Fa; Tang, Xiao Ying

    2018-03-01

    The manufacture of pressure vessels with austenitic stainless steel strain strengthening technology has become an important technical means for the light weight of cryogenic pressure vessels. In the process of increasing the strength of austenitic stainless steel, strain can induce the martensitic phase transformation in austenite phase. There is a quantitative relationship between the transformation quantity of martensitic phase and the basic mechanical properties. Then, the martensitic phase variables can be obtained by means of detection, and the mechanical properties and safety performance are evaluated and calculated. Based on this, the quantitative relationship between strain hardening and deformation induced martensite phase content is studied in this paper, and the mechanism of deformation induced martensitic transformation of austenitic stainless steel is detailed.

  2. A study of the mechanisms for the irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Solt, G.; Zimmermann, U.; Waeber, W.B.; Mercier, O.; Frisius, F.; Ghazi-Wakili, K.

    1987-03-01

    Irradiation damage particles were detected by small angle neutron scattering and positron annihilation techniques in two RPV steels. The particle radii were 8A and 14A prior to heat treatments for the plate and weldment, respectively; annealing leads to coarsening in the weldment, the volume fraction remains essentially constant at about 0.14%. The model of copper-rich precipitates 'diluted' by Mn atoms or, alternatively, by vacancy agglomerates is consistent with the neutron scattering data, the presence of simple voids in the weldment would contradict the positron results. Preliminary results on these steels and also on related alloys by methods 'new' in this field are reported. (author)

  3. 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

  4. 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.)

  5. Correlation of mechanical property changes in neutron-irradiated pressure vessel steels on the basis of spectral effects

    International Nuclear Information System (INIS)

    Heinisch, H.L.

    1991-01-01

    Comparisons are made of tensile data on specimens of A212B and A302B pressure vessel steels irradiated at low temperatures (40-90degC) and to low doses (<0.1 dpa) with 14 MeV D-T fusion neutrons in the Rotating Target Neutron Source (RTNS-II), with fission reactor neutrons in the Omega West Reactor (OWR) and the Oak Ridge Research Reactor (ORR), and with the highly thermal spectrum at the pressure vessel surveillance positions of the High Flux Isotope Reactor (HFIR). For each neutron spectrum, damage cross sections are determined for several defect production functions derived from atomistic computer simulations of collision cascades. Displacements per atom (dpa) and the numbers of freely migrating defects are tested as damage correlation parameters for the tensile data. The data from RTNS-II, OWR and ORR correlate fairly well when compared on the basis of dpa, but the data from HFIR show only about one sixth as many dpa are needed to produce the same radiation-induced yield stress changes as in the other neutron spectra. In the HFIR surveillance position a significant fraction of the displacements is produced by recoils resulting from thermal neutron captures. Having energies of about 400 eV, these recoils are much more efficient per unit energy at producing freely migrating defects than the high energy recoils responsible for most of the displacements in the other neutron spectra considered. A significantly better correlation of data from HFIR with those from the other spectra is achieved when the property changes are compared on the basis of the production of freely migrating self-interstitial defects. (orig./MM)

  6. SNPDL work in support of collaborative programmes to study fatigue crack propagation in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    McMinn, A.

    1981-07-01

    Work performed at Springfields Nuclear Power Development Laboratories (SNL) in support of United Kingdom and international cooperative groups is described. For the UK collaborative group crack growth tests have been performed on A533-B pressure vessel steel in air at 1 Hz, and in a simulated PWR environment at 1 Hz and 0.0167 Hz, all tests being at a stress ratio (R) of 0.7. Tests were performed at SNL at ambient temperature and pressure. The results showed that enhancements in crack growth rates were obtained because of the aqueous environment and the lower cyclic frequency. Good inter-laboratory agreement was obtained for the air test, indicating that there was little variation in the mechanical control between laboratories. There was also good agreement between the results from laboratories where tests were performed under high temperature wet conditions at 1 Hz. SNL room temperature data demonstrated an effect of temperature, as the crack growth rates found were up to a factor of four lower than the high temperature data. Although crack growth could be maintained at 0.0167 Hz and low cyclic stress intensity levels at ambient temperature, this was not found to be possible by the laboratories performing the test at 288 0 C. The first international 'round robin' test at R = 0.2 and 0.0167 Hz (1 cycle/min) in an aqueous environment gave a significant amount of scatter in the results. (author)

  7. Small specimen measurements of dynamic fracture toughness of heavy section steels for nuclear pressure vessel

    International Nuclear Information System (INIS)

    Tanaka, Y.; Iwadate, T.; Suzuki, K.

    1987-01-01

    This study presents the dynamic fracture toughness properties (KId) of 12 heats of RPV steels measured using small specimens and analysed based on the current research. The correlation between the KId test and other engineering small specimen tests such as Charpy test and drop weight test are also discussed and a method to predict the KId value is presented. (orig./HP)

  8. Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M [AB Atomenergi, Nykoeping (Sweden); Myers, H P [Chalmers Institute of Technology, Goeteborg (Sweden); Hannerz, N E [Motala Verkstads AB, Motala (Sweden)

    1967-09-15

    Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10{sup 18} and 5.6 x 10{sup 18} n/cm{sup 2} (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature.

  9. Dislocation structures in 16MND5 pressure vessel steel strained in uniaxial tension

    Czech Academy of Sciences Publication Activity Database

    Obrtlík, Karel; Robertson, Ch.; Marini, B.

    2005-01-01

    Roč. 342, - (2005), s. 35-41 ISSN 0022-3115 R&D Projects: GA AV ČR(CZ) 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : bainitic steels * dislocation structure * low temperature deformation Subject RIV: JG - Metallurgy Impact factor: 1.414, year: 2005

  10. Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Grounes, M.; Myers, H.P.; Hannerz, N.E.

    1967-09-01

    Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10 18 and 5.6 x 10 18 n/cm 2 (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature

  11. Hot Deformation Behavior of SA508Gr.4N Steel for Reactor Pressure Vessels

    Directory of Open Access Journals (Sweden)

    YANG Zhi-qiang

    2017-08-01

    Full Text Available The high-temperature plastic deformation and dynamic recrystallization behavior of SA508Gr.4N steel were investigated through hot deformation tests in a Gleeble1500D thermal mechanical simulator. The compression tests were performed in the temperature range of 1050-1250℃ and the strain rate range of 0.001-0.1s-1 with true strain of 0.16. The results show that from the high-temperature true stress-strain curves of the SA508Gr.4N steel, the main feature is dynamic recrystallization,and the peak stress increases with the decrease of deformation temperature or the increase of strain rate, indicating the experimental steel is temperature and strain rate sensitive material. The constitutive equation for SA508Gr.4N steel is established on the basis of the true stress-strain curves, and exhibits the characteristics of the high-temperature flow behavior quite well, while the activation energy of the steel is determined to be 383.862kJ/mol. Furthermore, an inflection point is found in the θ-σ curve, while the -dθ/dσ-σ curve shows a minimum value. The critical strain increases with increasing strain rate and decreasing deformation temperature. A linear relationship between critical strain (εc and peak strain (εp is found and could be expressed as εc/εp=0.517. The predicted model of critical strain could be described as εc=8.57×10-4Z0.148.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sun Mingyue, E-mail: mysun@imr.ac.cn [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China); Luhan, Hao; Shijian, Li; Dianzhong, Li; Yiyi, Li [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China)

    2011-11-15

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

  13. The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, R.B.; Buswell, J.T.

    1987-02-01

    An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. The failure behavior of duplex 316 L steel-TA6V titanium alloy spherical pressure vessels

    International Nuclear Information System (INIS)

    Miannay, D.

    1980-05-01

    The purpose of this paper is to compare the experimental residual stresses of spherical vessels made of TA6V alloy which exhibits plasticity before failure in toughness testing and cracked with several configurations, with stresses estimated according to the afore mentioned theories. An internal austenitic 316 L steel is used to prevent 'leak before break' [fr

  16. Thermal embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

    1995-01-01

    As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

  17. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  18. Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.P.; Viehrig, H.W.

    1992-01-01

    This work deals with the mechanical properties of RPV steels used WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy V-notch specimens and tensile test specimens were irradiated in the WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 x 10 18 and 5 x 10 19 n/cm 2 (E>1MeV). Post-irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. (orig.)

  19. Hydrogen Absorption Induced Slow Crack Growth in Austenitic Stainless Steels for Petrochemical Pressure Vessel Industries

    Directory of Open Access Journals (Sweden)

    Ronnie Rusli

    2011-05-01

    Full Text Available Type 304Land type 309 austenitic stainless steels were tested either by exposed to gaseous hydrogen or undergoing polarized cathodic charging. Slow crack growth by straining was observed in type 304L, and the formation of α‘ martensite was indicated to be precursor for such cracking. Gross plastic deformation was observed at the tip of the notch, and a single crack grew slowly from this region in a direction approximately perpendicular to the tensile axis. Martensite formation is not a necessary condition for hydrogen embrittlement in the austenitic phase.

  20. Effects of commercial cladding on the fracture behavior of pressure vessel steel plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Alexander, D.J.; Bolt, S.E.; Cook, K.V.; Corwin, W.R.; Oland, B.C.; Nanstad, R.K.; Robinson, G.C.

    1988-01-01

    The objective of this program is to determine the effect, if any, of stainless steel cladding upon the propagation of small surface cracks subjected to stress states similar to those produced by thermal shock conditions. Preliminary results from testing at temperature 10 deg. C and 60 deg. C below NDT have shown that (1) a tough surface layer (cladding and/or HAZ) has arrested running flaws under conditions where unclad plates have ruptured, and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate. (author)

  1. Survey of welding processes for field fabrication of 2 1/4 Cr-1 Mo steel pressure vessels. [128 references

    Energy Technology Data Exchange (ETDEWEB)

    Grotke, G.E.

    1980-04-01

    Any evaluation of fabrication methods for massive pressure vessels must consider several welding processes with potential for heavy-section applications. These include submerged-arc and shielded metal-arc, narrow-joint modifications of inert-gas metal-arc and inert-gas tungsten-arc processes, electroslag, and electron beam. The advantage and disadvantages of each are discussed. Electroslag welding can be dropped from consideration for joining of 2 1/4 Cr-1 Mo steel because welds made with this method do not provide the required mechanical properties in the welded and stress relieved condition. The extension of electron-beam welding to sections as thick as 4 or 8 inches (100 or 200 mm) is too recent a development to permit full evaluation. The manual shielded metal-arc and submerged-arc welding processes have both been employed, often together, for field fabrication of large vessels. They have the historical advantage of successful application but present other disadvantages that make them otherwise less attractive. The manual shielded metal-arc process can be used for all-position welding. It is however, a slow and expensive technique for joining heavy sections, requires large amounts of skilled labor that is in critically short supply, and introduces a high incidence of weld repairs. Automatic submerged-arc welding has been employed in many critical applications and for welding in the flat position is free of most of the criticism that can be leveled at the shielded metal-arc process. Specialized techniques have been developed for horizontal and vertical position welding but, used in this manner, the applications are limited and the cost advantage of the process is lost.

  2. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-01-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

  3. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

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

  4. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kocik, J.; Keilova, E.

    1993-01-01

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  5. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

  6. Topical problems of crackability in weld annealing of low-alloyed pressure vessel steels

    International Nuclear Information System (INIS)

    Holy, M.

    1977-01-01

    The following method was developed for determining annealing crackability: A sharp notch was made in the middle of the bodies of rods imitated in a welding simulator. Chucking heads were modified such as to permit chucking a rod in an austenitic block by securing the nut. Prestress was controlled by button-headed screw adapters. The blocks were made of 4 types of austenitic steels with graded thermal expansivity coefficients, all higher than that of the tested low-alloyed steel rod. The blocks with rods were placed in a furnace and heated at a rate of 100 degC/h. As a result of the larger austenite block diameter the rod began to be stretched and at some temperature of more than 500 degC it was pulled apart. The risk of annealing crackability of welded joints may be reduced by the choice of material and melt and by the technology of welding, mainly by the choice of a suitable addition material in whose weld metal the plastic deformation preferably takes place in annealing. (J.P.)

  7. Fracture toughness evaluation in the transition region of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Onizawa, K.; Suzuki, M.

    1995-01-01

    The fracture toughness (K jc and Jc) values at the cleavage fracture initiation in the transition region of a RPV steel were investigated using mainly precracked Charpy specimens. A conventional statistical approach and a fractographic study were applied to analyze the scatter of the fracture toughness values from precracked Charpy specimens. The material used was an ASTM A533B class 1 steel, which was designated as an IAEA correlation monitor material, JRQ. A lower bound transition curve of the fracture toughness for unirradiated condition was determined by the 5% confidence limit from the Weibull and fractographic analyses. The lower bound transition curve after irradiation was evaluated based on the statistics of unirradiated specimens. The results indicated that the shift of the fracture toughness transition curbe were somewhat larger than the Charpy 41J transition temperature. The parameters to determine the lower bound toughness such as the Weibull slope and the amount of ductile crack growth are discussed. The results are also compared with a model based on weakest link theory. (author). 12 refs, 12 figs, 5 tabs

  8. Radiation annealing mechanisms of low-alloy reactor pressure vessel steels dependent on irradiation temperature and neutron fluence

    International Nuclear Information System (INIS)

    Pachur, D.

    1982-01-01

    Heat treatment after irradiation of reactor pressure vessel steels showed annealing of irradiation embrittlement. Depending on the irradiation temperature, the embrittlement started to anneal at about 220 0 C and was completely annealed at 500 0 C with 4 h of annealing time. The annealing behavior was normally measured in terms of the Vickers hardness increase produced by irradiation relative to the initial hardness as a function of the annealing temperature. Annealing results of other mechanical properties correspond to hardness results. During annealing, various recovery mechanisms occur in different temperature ranges. These are characterized by activation energies from 1.5 to 2.1 eV. The individual mechanisms were determined by the different time dependencies at various temperatures. The relative contributions of the mechanisms showed a neutron fluence dependence, with the lower activation energy mechanisms being predominant at low fluence and vice versa. In the temperature range where partial annealing of a mechanism took place during irradiation, an increase in activation energy was observed. Trend curves for the increase in transition temperature with irradiation, for the relative increase of Vickers hardness and yield strength, and for the relative decrease of Charpy-V upper shelf energy are interpreted by the behavior of different mechanisms

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

    Science.gov (United States)

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

    2012-06-01

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the δ-ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the δ-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

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

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T.; Matsukawa, Y. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2012-06-15

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the {delta}-ferrite phase but not in the austenitic phase. Thermal aging at 400 Degree-Sign C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the {delta}-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the {gamma}-austenite and {delta}-ferrite interface. There were no Cr depleted zones around the carbide.

  11. On the dynamic fracture toughness and crack tip strain behavior of nuclear pressure vessel steel: Application of electromagnetic force

    International Nuclear Information System (INIS)

    Yagawa, G.; Yoshimura, S.

    1986-01-01

    This paper is concerned with the application of the electromagnetic force to the determination of the dynamic fracture toughness of materials. Taken is an edge-cracked specimen which carries a transient electric current and is simply supported in a steady magnetic field. As a result of their interaction, the dynamic electromagnetic force occurs in the whole body of the specimen, which is then deformed to fracture in the opening mode of cracking. Using the electric potential and the J-R curve methods to determine the dynamic crack initiation point in the experiment, together with the finite element method to calculate the extended J-integral with the effects of the electromagnetic force and inertia, the dynamic fracture toughness values of nuclear pressure vessel steel A508 class 3 are evaluated over a wide temperature range from lower to upper shelves. The strain distribution near the crack tip in the dynamic process of fracture is also obtained by applying a computer picture processing. (orig.)

  12. A study of the mechanical property changes of irradiation embrittled pressure vessel steels and their response to annealing treatments

    International Nuclear Information System (INIS)

    Tipping, P.; Waeber, W.B.; Mercier, O.

    1991-01-01

    Isochronal and isothermal heat treatments have been used to study the recovery of hardness of a neutron irradiated pressure vessel steel forging for the purposes of planning and realizing IAR (Irradiated-Annealed-Reirradiated) experiments. Charpy V notch tests have been performed to assess the toughness of the material irradiated to various fluences up to a maximum of 5 x 10 19 n/cm 2 , E>1 MeV at 290 o C with and without an intermediate annealing treatment at 450 o C x 168 h. The effect of the intermediate annealing was evident. The recovery of the upper shelf energies was strongly enhanced by a thermal ageing effect due to the annealing treatment for all fluence levels investigated compared to the irradiated condition. The transition temperature shifts exhibited a less straightforward behaviour due to the mentioned ageing effect which opposed the recovery process for this property leading to a net shift increase at lower and to a net recovery benefit at higher fluence levels. A phenomenological model description for the IAR embrittlement-recovery path is suggested. For this material and these irradiation conditions a plant life extension (PLEX) may be brought about if a specific annealing treatment is applied at a fluence level that is half the anticipated target fluence F for PLEX. In this case it was found that F>1.6 x 10 19 n/cm 2 . (author)

  13. Effects of degradation on the mechanical properties and fracture toughness of a steel pressure-vessel weld metal

    International Nuclear Information System (INIS)

    Wu, S.J.; Knott, J.F.

    2003-01-01

    A degradation procedure has been devised to simulate the effect of neutron irradiation on the mechanical properties of a steel pressure-vessel weld metal. The procedure combines the application of cold prestrain together with an embrittling heat treatment to produce an increase in yield stress, a decrease in strain hardening rate, and an increased propensity for brittle intergranular fracture. Fracture tests were carried out using blunt-notch four-point-bend specimens in slow bend over a range of temperatures and the brittle/ductile transition was shown to increase by approximately 110 deg. C as a result of the degradation. Fractographic analysis of specimens broken at low temperatures showed about 30% intergranular failure in combination with transgranular cleavage. Predictions have been made of the ductile-brittle transition curves for the weld metal (sharp crack) fracture toughness in degraded and non-degraded states, based on the notched-bar test results and on finite element analyses of the stress distributions ahead of the notches and sharp cracks. The ductile-brittle transition temperature shift (ΔT=110 deg. C) between non-degraded and degraded weld metal at a notch opening displacement of 0.31 mm was combined with the Ritchie, Knott and Rice (RKR) model to predict an equivalent shift of 115 deg. C for sharp-crack specimens at a toughness level of 70 MN/m 3/2

  14. Estimation of lower-bound KJc on pressure vessel steels from invalid data

    International Nuclear Information System (INIS)

    McCable, D.E.; Merkle, J.G.

    1996-01-01

    Statistical methods are currently being introduced into the transition temperature characterization of ferritic steels. Objective is to replace imprecise correlations between empirical impact test methods and universal K Ic or K Ia lower-bound curves with direct use of material-specific fracture mechanics data. This paper introduces a computational procedure that couples order statistics, weakest-link statistical theory, and a constraint model to arrive at estimates of lower-bound K Jc values. All of the above concepts have been used before to meet various objectives. In the present case, scheme is to make a best estimate of lower-bound fracture toughness when resource K Jc data are too few to use conventional statistical analyses. Utility of the procedure is of greatest value in the middle-to-high toughness part of the transition range where specimen constraint loss and elevated lower-bound toughness interfere with conventional statistical analysis methods

  15. Assessment of models predicting irradiation effects on tensile properties of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Pineau, L.; Landron, C.

    2015-01-01

    In this paper, an analysis of tensile data acquired as part of the French Reactor Vessel Surveillance Program (RVSP) is produced. This program contains amongst other mechanical tests, tensile tests at 20 and 300 C degrees on non irradiated base metals and at 300 C degrees only on irradiated materials. It shows that irradiation leads to an increase in the yield strength and a decrease in the strain hardening. The exploitation of tensile results has permitted to express a relationship between yield strength increase measured and fluence value, as well as between strain hardening decrease and yield strength evolution. The use of these relations in the aim at predicting evolution of tensile properties with irradiation has then permitted to propose a methodology to model entire stress-strain curves of irradiated base metal only based on the non irradiated stress-strain curve. These predictions were successfully compared with an experimental standard case. (authors)

  16. Ductile fracture toughness of heavy section pressure vessel steel plate. A specimen-size study of ASTM A 533 steels

    International Nuclear Information System (INIS)

    Williams, J.A.

    1979-09-01

    The ductile fracture toughness, J/sub Ic/, of ASTM A 533, Grade B, Class 1 and ASTM A 533, heat treated to simulate irradiation, was determined for 10- to 100-mm thick compact specimens. The toughness at maximum specimen load was also measured to determine the conservatism of J/sub Ic/. The toughness of ASTM A 533, Grade B, Class 1 steel was 349 kJ/m 2 and at the equivalent upper shelf temperature, the heat treated material exhibited 87 kJ/m 2 . The maximum load fracture toughness was found to be linearly proportional to specimen size, and only specimens which failed to meet ASTM size criteria exhibited maximum load toughness less than J/sub Ic/

  17. Stress induced transformation to bainite in Fe-Cr-Mo-C pressure vessel steel

    International Nuclear Information System (INIS)

    Bhadeshia, H.K.D.H.; David, S.A.; Vitek, J.M.; Reed, R.W.

    1991-01-01

    The kinetics of the bainitic transformation in a polycrystalline Fe-Cr-Mo-C alloy designed for applications in energy generation systems has been studied, with particular attention to the influence of mild tensile stresses on transformation behaviour. The steel was found to exhibit the incomplete reaction phenomenon, in which transformation to bainite stops well before the residual austenite acquires its paraequilibrium carbon concentration. It was found that even in the absence of an applied stress, the growth of bainitic ferrite caused anisotropic changes in specimen dimensions, consistent with the existence of crystallographic texture in its austenitic condition and, significantly, with the nature of the invariant-plane strain shape change that accompanies the growth of bainitic ferrite. Thus, transformation induced plasticity could be detected in fine grained polycrystalline samples, even in the absence of applied stress. The application of an external stress was found to alter radically the transformation behaviour, with clear evidence that the stress tends to favour the development of certain crystallographic variants of bainite, even though the stress may be well below the single phase yield strength. It is concluded that the transformation is influenced significantly by stresses as low as 45 MN m -2 , even though the effect may not be obvious in metallographic studies. The results are analysed and discussed in terms of the mechanism of the bainite transformation. (author)

  18. Size effects on failure behaviour of reactor pressure vessel steel and their dependence on deformation inhomogeneity

    International Nuclear Information System (INIS)

    Aktaa, J.; Klotz, M.; Schmitt, R.

    2003-01-01

    The investigation of the size dependence of the material behaviour and particularly of the failure strain is the main objective of the European research project LISSAC (Limit Strains for Severe Accident Conditions). Within our activities in LISSAC, tensile test series with specimens of similar geometry and different sizes are performed. The specimens, cut from the wall of a real reactor vessel, are flat with a central hole, flat with a double edge notch as well as round with a circumferential notch in order to obtain inhomogeneous deformation with high strain gradients, which will be higher in the smaller specimens and might be responsible for size effects. An additional variation of the strain gradient is obtained by varying the central hole radius of the flat specimens, with three different hole geometries being considered: round hole, increased round hole and slot. During the tests optical methods are used for measuring local deformations and partly local strain gradients. The results obtained show a size effect neither on the global nor on the local deformation behaviour, whereas the damage and failure behaviour is influenced significantly by the size of the specimen. On the basis of the surface deformation measurements, finite element calculations are performed to estimate the local failure strains as well as the corresponding strain gradients. A clear dependence of local failure strains on strain gradients is obtained. (author)

  19. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels; Evaluacion de los defectos inducidos por la radiacion neutronica en los aceros de vasijas

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J

    1978-07-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs.

  20. Thermodynamic Alloy Design of High Strength and Toughness in 300 mm Thick Pressure Vessel Wall of 1.25Cr-0.5Mo Steel

    Directory of Open Access Journals (Sweden)

    Hye-sung Na

    2018-01-01

    Full Text Available In the 21st century, there is an increasing need for high-capacity, high-efficiency, and environmentally friendly power generation systems. The environmentally friendly integrated gasification combined-cycle (IGCC technology has received particular attention. IGCC pressure vessels require a high-temperature strength and creep strength exceeding those of existing pressure vessels because the operating temperature of the reactor is increased for improved capacity and efficiency. Therefore, high-pressure vessels with thicker walls than those in existing pressure vessels (≤200 mm must be designed. The primary focus of this research is the development of an IGCC pressure vessel with a fully bainitic structure in the middle portion of the 300 mm thick Cr-Mo steel walls. For this purpose, the effects of the alloy content and cooling rates on the ferrite precipitation and phase transformation behaviors were investigated using JMatPro modeling and thermodynamic calculation; the results were then optimized. Candidate alloys from the simulated results were tested experimentally.

  1. 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

  2. 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)

  3. Review of recent studies on neutron irradiation embrittlement in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Sudo, Akira; Miyazono, Shohachiro

    1983-06-01

    Recent studies in foreign countries (USA, France, FRG and UK) on neutron irradiation embrittlement have been reviewed. These studies are classified into four areas, such as 1) effect of chemical composition on irradiation embrittlement sensitivity, 2) postirradiation heat treatment for embrittlement relief, 3) fracture toughness evaluation of irradiated materials based on fracture mechanics analysis, and 4) effect of irradiation on fatigue crack propagation behavior. The first area mainly includes the studies related to the effects of copper, phosphorus impurities and nickel alloying and synergistic effect of these components, and furthermore, evaluation of Regulatory Guide 1.99 Rev.l. Studies in the second area show the effects of annealing condition (temperature and time) and metallugical condition on embrittlement relief, and evaluation of periodic annealing in the period of irradiation as a promising method for embrittlement control. Studies in the third area show the correlation between fracture toughness and Cv notch ductility changes with neutron irradiation, and J-R curves of irradiated materials based on the elasto-plastic fracture mechanics. In the forth area, most of studies are investigated in air condition but a few studies in reactor-grade water at high temperature and pressure. (author)

  4. Low Temperature (320 deg C and 340 deg C) Creep Crack Growth in Low Alloy Reactor Pressure Vessel Steel

    International Nuclear Information System (INIS)

    Rui Wu; Sandstroem, Rolf; Seitisleam, Facredin

    2004-02-01

    Uni-axial creep and creep crack growth (CCG) tests at 320 deg C and 340 deg C as well as post test metallography have been carried out in a low alloy reactor pressure vessel steel (ASTM A508 class 2) having simulated coarse grained heat affected zone microstructure. The CCG behaviour is studied in terms of steady crack growth rate, creep fracture parameter C*, stress intensity factor and reference stress at given testing conditions. It has been found that CCG does occur at both tested temperatures. The lifetimes for the CCG tests are considerably shorter than those for the uni-axial creep tests. This is more pronounced at longer lifetimes or lower stresses. Increasing temperature from 320 deg C to 340 deg C causes a reduction of lifetime by approximately a factor of five and a corresponding increase of steady crack growth rate. For the CCG tests, there are three regions when the crack length is plotted against time. After incubation, the crack grows steadily until it accelerates when rupture is approached. Notable crack growth takes place at later stage of the tests. No creep cavitation is observed and transgranular fracture is dominant for the uni-axial creep specimens. In the CT specimens the cracks propagate intergranularly, independent of temperature and time. Some relations between time to failure, reference stress and steady crack growth rate are found for the CCG tests. A linear extrapolation based on the stress-time results indicates that the reference stress causing failure due to CCG at a given lifetime of 350,000 hours at 320 deg C is clearly lower than both yield and tensile strengths, on which the design stress may have based. Therefore, caution must be taken to prevent premature failure due to low temperature CCG. Both uni-axial and CCG tests on real welded joint at 320 deg C, study of creep damage zone at crack tip as well as numerical simulation are recommended for future work

  5. The effect of specimen size on the ductile/brittle transition temperature in an A533B pressure vessel steel

    International Nuclear Information System (INIS)

    Green, G.; Knott, J.F.

    It was ascertained that it is possible to relate critical crack opening displacement (COD) values, deltasub(crit), obtained on small specimens of A 533-B pressure vessel steel to the fracture toughness value representing the initiation of fracture in a large structure. The variation of deltasub(crit) with temperature is given. A sharp increase in deltasub(crit) is observed above a temperature of approximately -100 degC and this was found to be associated with the initiation of small amounts of fibrous fracture, prior to a cleavage instability. An upper limit to the deltasub(crit) values was obtained above -50 degC, where the fracture was found to be fully ductile. Values of deltasub(crit) estimated from the valid fracture toughness results are shown for comparison. At low temperatures the estimated deltasub(crit) values are seen to be less than those measured in the small bend specimens and the sharp increase in the estimated deltasub(crit) values occurs at a higher temperature, approximately 0 degC. The room temperature deltasub(crit) value, estimated from the valid toughness results (0.15 mm) compares well with COD for the initiation of fibrous fracture, measured at the same temperature in small bend specimens (0.175 mm). The following conclusions were drawn from the experiments: 1. The ductile/brittle transition temperature, determined by critical COD measurements, is influenced by the relaxation of triaxial stresses in small specimens. 2. It is possible to relate critical COD values for the initiation of fibrous fracture, measured in small specimens, to the fracture toughness representing this behaviour in a large structure

  6. Application of ductile fracture assessment methods for the assessment of pressure vessels from high strength steels (HSS)

    International Nuclear Information System (INIS)

    Eisele, U.; Schiedermaier, J.

    2003-01-01

    The economical and safe design of pressure vessels requires, besides others, also a detailed knowledge of the vessel failure behaviour in the case of existing imperfections or cracks. The behaviour of a cracked component under a given loading situation depends on material toughness. For ferritic steels, the material toughness is varying with temperature. At low temperature dominantly brittle fracture behaviour is observed, at high temperature the failure mode is dominantly ductile fracture. The transition between these two extremes is floating. In the case of existing or postulated cracks, the safety analysis has to be performed using fracture mechanics methods. In the lower shelf of toughness, K iC as of ASTM E 399 is the characterising value for crack initiation and immediate unstable crack extension (cleavage). In the upper shelf level the characterising value is the ''actual crack initiation toughness'' J i acc. to ISO 12135, characterising the onset of slow stable crack extension. For the transition regime in ASTM E 1921 the instability values K JC are defined, characterising cleavage failure after more or less extended ductile crack growth. The safety analysis of a component operated in the upper shelf of the material toughness, has to consider initiation as well as stable crack extension following initiation. The inclusion of any crack extension into this consideration needs to consider the influence of the constraint in front of a crack tip, leading to multiaxial stress conditions and decreasing the material crack resistance significantly. Thus, the exclusion of crack initiation needs to be proven in a first step of each safety analysis. Assessing the component in a uniform way over the relevant temperature range is possible by using initiation characteristics, which also have the advantage of transferability. A change of criterion considering initiation at the lower shelf, instability in the transition range and again initiation in the upper shelf can be

  7. Changes of the mechanical properties of ASTM A 533 type B class 1 (JRQ) steel used in pressure vessels of nuclear power plants

    International Nuclear Information System (INIS)

    Balderrama, Juan J.; Iorio, Antonio F.

    1999-01-01

    The steels used in pressure vessels generally present a non-homogenous microstructure across the thickness of their walls due to their manufacturing process. Average thickness being between 200-250 mm also makes the problem more serious. These facts lead us to think that the variation affects not only microstructure, but also mechanical properties. For this reason the methodology for the evaluation of materials should be standardized for their use before and after radiation by means of a surveillance program which allows us to verify the conditions of the steel of the pressure vessel by using Charpy-v, tensile and fracto-mechanics specimens inside the reactor to obtain information about the condition of the pressure vessel material. In order to analyze these changes, tests were carried out using Charpy-v specimens with different orientation inside the block representing the wall thickness and the corresponding ductile-to-brittle transition curves were made for each direction. The orientations to be considered will be four in all and will be those called TL, LT, ST and LS by ASTM E 399 (1993). The conclusions reached arise from a comparative analysis of the results obtained for each orientation under study and confirm the recommendation by Standards regarding the selection of the TL orientation as the most conservative. (author)

  8. Assuring reliability of unconventional weld joint configurations in austenitic stainless steel pressure vessels through non-destructive examination

    International Nuclear Information System (INIS)

    Jayakumar, I.; Manimohan, M.; Chandrasekaran, G.V.; Abdul Majeeth, S.; Subrahmanyam, P.S.

    1996-01-01

    Design of weld configurations in engineering structures is based on NDE inspectability apart from other considerations. They are mostly standardised. This paper deals with the development of an effective NDE methodology for an unconventional weld joint configuration occurring in a critical pressure vessel with edge preparation orientations different from that normally encountered in fabrication of such vessels. It is K-type butt joint between a heavy load bearing member and a curved vessel wall resulting in an oblique fillet weld. The heavy load bearing functional requirement needs a high integrity fail safe joint during its operating life and the stringent quality level specified by customer was ensured at every stage of its workmanship through effective NDE relying on conventional methods as explained. (author)

  9. Evaluation of creep damage due to stress relaxation in SA533 grade B class 1 and SA508 class 3 pressure vessel steels

    International Nuclear Information System (INIS)

    Hoffmann, C.L.; Urko, W.

    1993-01-01

    Creep damage can result from stress relaxation of residual stresses in components when exposed to high temperature thermal cycles. Pressure vessels, such as the reactor vessel of the modular high-temperature gas reactor (MHTGR), which normally operate at temperatures well below the creep range can develop relatively high residual stresses in high stress locations. During short term excursions to elevated-temperatures, creep damage can be produced by the loadings on the vessel. In addition, residual stresses will relax out, causing greater creep damage in the pressure vessel material than might otherwise be calculated. The evaluation described in this paper assesses the magnitude of the creep damage due to relaxation of residual stresses resulting from short term exposure of the pressure vessel material to temperatures in the creep range. Creep relaxation curves were generated for SA533 Grade B, Class 1 and SA508 Class 3 pressure vessel steels using finite element analysis of a simple uniaxial truss loaded under constant strain conditions to produce an initial axial stress equal to 1.25 times the material yield strength at temperature. The strain is held constant for 1000 hours at prescribed temperatures from 700 F to 1000 F. The material creep law is used to calculate the relaxed stress for each time increment. The calculated stress relaxation versus time curves are compared with stress relaxation test data. Creep damage fractions are calculated by integrating the stress relaxation versus time curves and performing a linear creep damage summation using the minimum stress to rupture curves at the respective relaxation temperatures. Cumulative creep damage due to stress relaxation as a function of time and temperature is derived from the linear damage summation

  10. Time-dependent temper embrittlement of reactor pressure vessel steel: Correlation between microstructural evolution and mechanical properties during tempering at 650 °C

    Energy Technology Data Exchange (ETDEWEB)

    Li, Chuanwei; Han, Lizhan; Yan, Guanghua; Liu, Qingdong; Luo, Xiaomeng; Gu, Jianfeng, E-mail: gujf@sjtu.edu.cn

    2016-11-15

    The microstructural evolution of reactor pressure vessel (RPV) steel and its effect on the mechanical properties during tempering at 650 °C were studied to reveal the time-dependent toughness and temper embrittlement. The results show that the toughening of the material should be attributed to the decomposition of the martensite/austenite constituents and uniform distribution of carbides. When the tempering duration was 5 h, the strength of the investigated steel decreased to strike a balance with the material impact toughness that reached a plateau. As the tempering duration was further increased, the material strength was slightly reduced but the material impact toughness deteriorated drastically. This time-dependent temper embrittlement is different from traditional temper embrittlement, and it can be partly attributed to the softening of the matrix and the broadening of the ferrite laths. Moreover, the dimensions and distribution of the grain carbides are the most important factors of the impact toughness. - Highlights: • The fracture mechanism of reactor pressure vessel (RPV) steels under impact load was investigated. • The Charpy V-notch impact test and the hinge model were employed for the study. • Grain boundary carbides play a key role in the impact toughness and fracture toughness. • The dependence of the deterioration of impact toughness on tempering time was analyzed for the first time.

  11. Biaxial Loading Tests for steel containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Miyagawa, T. [Nuclear Power Engineering Corp., Tokyo (Japan); Wright, D.J.; Arai, S.

    1999-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  12. Biaxial Loading Tests for steel containment vessel

    International Nuclear Information System (INIS)

    Miyagawa, T.; Wright, D.J.; Arai, S.

    1999-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  13. 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

  14. A study of the pressure vessel steel of the WWER-440 unit 1 of the Kozloduy nuclear power plant

    Science.gov (United States)

    Kostadinova, E.; Velinov, N.; Avdjieva, T.; Mitov, I.; Rusanov, V.

    2017-11-01

    A comparison between highly neutron irradiated samples from the region of weld № 4 and low irradiated samples from weld № 1 taken from the pressure vessel of the WWER-440 Unit № 1 of the Kozloduy NPP has been performed. Measurements of the residual activity of samples from the outer surface of the reactor pressure vessel bottom corpus reveal very low activity of 60Co. Insofar as there the base and weld metal appear to be exposed to a very low neutron fluence, the samples from these locations can be considered as practically not affected and may serve as a reference basis for comparison with highly irradiated pressure vessel regions. The Mössbauer parameters isomer shift (IS) and quadrupole splitting (QS) were found to be absolutely irradiation insensitive. A stepwise reduction of the internal hyperfine magnetic field Bhf, each by about 2.6 T, was observed. This can be attributed to the replacement of one or two surrounding iron atoms as first nearest neighbors by non-iron alloying atoms. The Mössbauer experimental line widths for irradiated and non-irradiated samples are practically the same, which is a quite unexpected result. The area fraction ratio for the three main Zeeman sextet subspectra S1:S2:S3 shows very high irradiation sensitivity. For the bottom low irradiated region of the reactor vessel the values are S1:S2:S3 = 50.1:40.0:9.4. After seven years of operation between the pressure vessel annealing in 1989 and the autumn of 1996 when the samples from weld № 4 were taken the ratio changes strongly to S1:S2:S3 = 56.4:34.7:8.5. A possible explanation of this result is that neutron irradiation gives rise to a precipitation process involving predominantly alloying atoms as Ni, Mn, Cr, Mo and V which become mobile and precipitate in the form of carbides and/or P-rich phases and alloying atom aggregates. This "refinement" process lowers the partial area of subspectra S2 and S3 where alloying atoms are involved and leads to a higher area fraction of

  15. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Science.gov (United States)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  16. 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

  17. 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.)

  18. 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

  19. Results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Luk, V.K.; Ludwigsen, J.S.; Hessheimer, M.F.; Komine, Kuniaki; Matsumoto, Tomoyuki; Costello, J.F.

    1998-05-01

    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 US Nuclear Regulatory Commission. Two tests are being conducted: (1) a test of a model of a steel containment vessel (SCV) and (2) a test of a model of a prestressed concrete containment vessel (PCCV). This paper summarizes the conduct of the high pressure pneumatic test of the SCV model and the results of that test. Results of this test are summarized and are compared with pretest predictions performed by the sponsoring organizations and others who participated in a blind pretest prediction effort. Questions raised by this comparison are identified and plans for posttest analysis are discussed

  20. 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

  1. Fracture mechanics analysis of reactor pressure vessel under pressurized thermal shock - The effect of elastic-plastic behavior and stainless steel cladding -

    International Nuclear Information System (INIS)

    Joo, Jae Hwang; Kang, Ki Ju; Jhung, Myung Jo

    2002-01-01

    Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). The PTS event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored

  2. 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.

  3. 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

  4. Unaxial stress relaxation and creep behaviour in weldments of the pressure vessel steel A533B between 600 and 640 degree C

    International Nuclear Information System (INIS)

    Otterberg, R.

    1979-10-01

    In order to predict the stress reduction during stress relief heat treatment in welded joints of the pressure vessel steel A533B, uniaxial stress relaxation as well as creep tests have been performed. The specimens were isothermally stress relaxed between 600 and 640 degree C from initial stresses corresponding to specimen elongations of 0.25, 0.5 and 0.2 percent. The stress relaxation results are excellently described by a Norton relationship. The magnitude of the initial stress has been found to affect the stress relaxation in the beginning of the tests, but at times longer than one hour the effect is very small. Creep strain data from creep tests in the actual temperature interval was converted to describe stress relaxation behaviour as well. The results will be used in a forthcoming study to predict the multiaxial stress reduction in thick weldments of A533B. (author)

  5. Evaluation of the a.c. potential drop method to determine J-crack resistance curves for a pressure vessel steel

    International Nuclear Information System (INIS)

    Gibson, G.P.

    1989-01-01

    An evaluation has been carried out of the a.c. potential drop technique for determining J-crack growth resistance curves for a pressure vessel steel. The technique involves passing an alternating current through the specimen and relating the changes in the potential drop across the crack mouth to changes in crack length occuring during the test. The factors investigated were the current and voltage probe positions, the a.c. frequency and the test temperature. In addition, by altering the heat treatment of the material, J-crack resistance curves were obtained under both contained and non-contained yielding conditions. In all situations, accurate J-R curves could be determined. (author)

  6. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects

    International Nuclear Information System (INIS)

    Meslin-Chiffon, E.

    2007-11-01

    The embrittlement of reactor pressure vessel (RPV) under irradiation is partly due to the formation of point defects (PD) and solute clusters. The aim of this work was to gain more insight into the formation mechanisms of solute clusters in low copper ([Cu] = 0.1 wt%) FeCu and FeCuMnNi model alloys, in a copper free FeMnNi model alloy and in a low copper French RPV steel (16MND5). These materials were neutron-irradiated around 300 C in a test reactor. Solute clusters were characterized by tomographic atom probe whereas PD clusters were simulated with a rate theory numerical code calibrated under cascade damage conditions using transmission electron microscopy analysis. The confrontation between experiments and simulation reveals that a heterogeneous irradiation-induced solute precipitation/segregation probably occurs on PD clusters. (author)

  7. The near-threshold high R-ratio fatigue crack growth characteristics of SA508 cl III reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Achilles, R.D.; Bulloch, J.H.

    1989-01-01

    This paper describes the effect of frequency and environment on the near-threshold fatigue crack growth behaviour of SA508 cl III reactor pressure vessel (RPV) steel. The study has shown that in the near-threshold regime microstructure and environment markedly affect fatigue crack growth behaviour. In an aqueous environment, fatigue crack growth behaviour became even more sensitive to microstructure, and the fatigue crack growth rate increased by a factor of four in the case of the 3Hz test, while that for the 0.3Hz test was increased by a factor of approximately sixteen. This environmental enhancement manifested itself in the form of intergranular failure. For the 0.3Hz test the percentage intergranular failure decreased from 18% to <1% with an increase in ΔK level. The transition from microstructure-sensitive to microstructure-insensitive occurs when the cyclic plastic zone size is of the order of the prior austenite grain size. (author)

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

    International Nuclear Information System (INIS)

    Voigt, J.

    1978-03-01

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

  9. Long term integrity of reactor pressure vessel and primary containment vessel after the severe accidents in Fukushima Daiichi Nuclear Power Station. Leaching property of spent oxide fuel segment and corrosion property of a carbon steel under artificial seawater immersion

    International Nuclear Information System (INIS)

    2014-06-01

    Primary containment vessel (PCV), reactor pressure vessel and pedestal in Fukushima Daiichi Nuclear power station units 1 through 3 have been exposed to severe thermal, chemical and mechanical conditions due to core meltdown events and seawater injections for emergent core cooling. These components will be immersed in diluted seawater with dissolved fission products under irradiation until the end of debris removal. Fresh water injected into the cores contacts with debris to cool, dissolves or erodes their constituents, mixed with retained water, and becomes 'accumulated water' with radioactive nuclides. We have focused the leaching of fission products into the accumulated water under lower temperature (323 K). FUGEN spent oxide fuel segments were immersed to determine the leaching factor of fission product and actinide elements. Since PCV made from carbon steel is one of the most important boundaries to prevent from fission products release, corrosion behavior has been paid attention to evaluate their integrity. Carbon steel specimens were immersion- and electrochemical-tested in diluted seawater with simulants of the accumulated water at 323 K in order to evaluate the effect of fission products in particular cesium and radiation. (author)

  10. 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)

  11. 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

  12. Low upper-shelf toughness, high transition temperature test insert in HSST [Heavy Section Steel Technology] PTSE-2 [Pressurized Thermal Shock Experiment-2] vessel and wide plate test specimens: Final report

    International Nuclear Information System (INIS)

    Domian, H.A.

    1987-02-01

    A piece of A387, Grade 22 Class 2 (2-1/4 Cr - 1 Mo) steel plate specially heat treated to produce low upper-shelf (LUS) toughness and high transition temperature was installed in the side wall of Heavy Section Steel Technology (HHST) vessel V-8. This vessel is to be tested by the Oak Ridge National Laboratory (ORNL) in the Pressurized Thermal Shock Experiment-2 (PTSE-2) project of the HSST program. Comparable pieces of the plate were made into six wide plate specimens and other samples. These samples underwent tensile tests, Charpy tests, and J-integral tests. The results of these tests are given in this report

  13. 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.

  14. Effect of chemical composition on irradiation embrittlement and annealing in Ni-Cr-Mo-V reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Novosad, P [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    Results concerning copper and phosphorus influence on radiation-induced changes in the Ni-Cr-Mo-V steel mechanical properties, are presented. Correlations between different mechanical properties for steels with different chemical composition, are presented. A comparison of transition temperature shifts obtained for static and dynamic fracture toughness tests and Charpy impact tests, is discussed. Recovery of radiation hardening, measured by hardness test after isochronal annealing of steels with different compositions, is also shown. Copper content strongly affects irradiation-induced changes of mechanical properties, but phosphorus content in connection with variable copper content has only a small effect. (author). 4 refs., 4 figs., 4 tabs.

  15. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany); Gillemot, F. [Centre for Energy Research of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Hernández-Mayoral, M.; Serrano, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Török, G. [Wigner Research Center for Physics of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Ulbricht, A.; Altstadt, E. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany)

    2015-06-15

    Highlights: • TEM and SANS were applied to estimate mean size and number density of loops, nanovoids and Cu-rich clusters. • A three-feature dispersed-barrier hardening model was applied to estimate the yield stress increase. • The values and errors of the dimensionless obstacle strength were estimated in a consistent way. • Nanovoids are stronger obstacles for dislocation glide than dislocation loops, loops are stronger than Cu-rich clusters. • For reactor-relevant conditions, Cu-rich clusters contribute most to hardening due to their high number density. - Abstract: Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  16. A study on the welding characteristics of Mn-Ni-Mo type A302-C steel plate for pressure vessel

    International Nuclear Information System (INIS)

    Yoon, Byoung Hyun; Chang, Woong Seong; Kweon, Young Gak

    2003-01-01

    In order to develop ASTM A302 grade C type steel plate with excellent weldability, several steels with different chemistry have been manufactured and evaluated their mechanical properties and weldability. Trial A302-C steels have revealed tensile strength in the range of 61-67kg/mm 2 and elongation in the range of 27∼32%, depending on chemical compositions within the ASTM specification range. In case of impact toughness, trial steels showed in the range of 58-70J at 0 .deg. C. From the weldability test, the minimum preheat temperature was found to be about 150 .deg. C, and automatic welding condition satisfied the requirements of both ASTM specification and users

  17. Prediction of cleavage crack propagation and arrest in a nuclear pressure vessel steel (16MND5) under thermal shock

    International Nuclear Information System (INIS)

    Yang, Xiaoyu

    2015-01-01

    The purpose of this PhD study is to predict the propagation and arrest of cleavage cracks in a French PWR vessel steel (16MND5). This is accomplished through use of a local criterion based on the critical stress calculated ahead of crack tip. Previous work has shown that fracture mechanism was cleavage associated with the ductile shear zone between the different planes of cracking. Thus, the critical stress at crack tip depends on stain rate. This thesis consists of numerical optimization, identification and validation of a local criterion based on experiments which have complex thermomechanical loads. The criterion accounts for various crack paths, deepening the knowledge about micro mechanisms during crack propagation in order to justify the established criterion. Criterion identification was carried out by using numerical simulations of tension tests performed on CT (Compact Tension) specimens at four different temperatures (-150 C, -125 C, -100 C and -75 C). The eXtended Finite Element Method (X-FEM) was used in CAST3M FE software to model dynamic crack propagation and arrest. The analysis results in 2D and 3D showed that the critical stress at crack tip increased with the inelastic strain rate. Therefore, a criterion based on the critical stress was established. An analytical model was developed to justify the identified criterion. The critical stress given by the local criterion was considered as the sum of the critical cleavage stress and the stress generated by the deformation of ligaments behind the crack tip. In order to quantify this phenomenon, measurements of ligaments' characteristics have been performed on fracture surfaces and on cross-sections of the specimens. The stress profile of the crack lips generated by ligaments was calculated by modeling of multi-cracks on specimen's cross-section. The contribution of stress generated by ligaments to the critical stress at crack tip was obtained with this method, and then the analytical model of

  18. Positron annihilation spectroscopy and small angle neutron scattering characterization of nanostructural features in high-nickel model reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Glade, Stephen C. [Nuclear Engineering Department, University of California, Berkeley, CA 94720-1730 (United States); Wirth, Brian D. [Nuclear Engineering Department, University of California, Berkeley, CA 94720-1730 (United States)]. E-mail: bdwirth@nuc.berkeley.edu; Odette, G. Robert [University of California, Santa Barbara, CA (United States); Asoka-Kumar, P. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2006-06-01

    Irradiation embrittlement in nuclear reactor pressure vessel steels results from the hardening by a high number density of nanometer scale features. In steels with more than {approx}0.10% Cu, the dominant features are often Cu-rich precipitates typically alloyed with Mn, Ni and Si. At low-Cu and low-to-intermediate Ni levels, so-called matrix hardening features are believed to be vacancy-solute cluster complexes, or their remnants. However, Mn-Ni-Si rich precipitates, with Mn plus Ni contents greater than Cu, can form at high alloy Ni contents and are promoted at irradiation temperatures lower than the nominal 290 deg. C. Even at very low-Cu levels, late blooming Mn-Ni-Si rich precipitates are a significant concern due to their potential to form large volume fractions of hardening features. Positron annihilation spectroscopy (PAS) and small angle neutron scattering neutron (SANS) measurements were used to characterize the fine-scale microstructure in split-melt A533B steels with varying Ni and Cu contents, irradiated at selected conditions from 270 to 310 deg. C between {approx}0.04 and 1.6 x 10{sup 23} n m{sup -2}. The objective was to assess the character, composition and magnetic properties of Cu-rich precipitates, as well as to gain insight on the matrix features. The results suggest that the irradiated very low-Cu and intermediate Ni steel contains small vacancy-Mn-Ni-Si cluster complexes, but not large, well-formed and highly enriched Mn-Ni-Si phases. The hardening features in steels containing 0.2% and 0.4% Cu, and 0.8% and 1.6% Ni are consistent with well-formed, non-magnetic Cu-Ni-Mn precipitates. The precipitate number densities and volume fractions increase, while their sizes decrease, with increasing Ni and decreasing irradiation temperature. The precipitates evolve with fluence in stages of nucleation, growth and limited coarsening.

  19. Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

  20. Evaluation of the magnetic and mechanical properties of reactor pressure vessel steels by incremental permeability change curve measurements

    International Nuclear Information System (INIS)

    Ebine, N.; Suzuki, M.

    2001-01-01

    Incremental permeability measurement was performed for two types of structural steels along with the magnetization of their hysteresis minor-loop. The obtained incremental permeability change curve has two sharp peaks, and the width between the two peaks is correlated with the coercivity. Hence the existence of good correlation was verified. On the basis of this result, nondestructive measurement experiments were carried out with planar coils to evaluate changes in the material properties of ferromagnetic structural steel plates. Changes in output voltages from planar coils with different test plates were correlated with their mechanical and magnetic properties. The correlation is so good that the measurement method adopted in this work could be used for nondestructive evaluation of material degradation in ferromagnetic structural steels. (author)

  1. 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.)

  2. 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.

  3. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

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

    1994-12-01

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

  4. Irradiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Bros, J.

    2000-01-01

    From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

  5. Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

    International Nuclear Information System (INIS)

    Fujii, K.; Fukuya, K.; Hojo, T.

    2013-01-01

    Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates

  6. Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, K., E-mail: fujiik@inss.co.jp [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Fukuya, K. [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Hojo, T. [Japan Nuclear Energy Safety Organization, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2013-11-15

    Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates.

  7. Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

    International Nuclear Information System (INIS)

    Whitman, G.D.

    1975-01-01

    The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

  8. 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)

  9. Crack behaviour of ferritic pressure vessels steels in oxygenated high temperature water under transient loadings. Crack corrosion phase 2. Crack development and fatigue. Final report

    International Nuclear Information System (INIS)

    Weissenberg, Thomas

    2014-03-01

    Using the example of the ferritic steels 22NiMoCr3-7 and 15MnNi6-3 representative for Nuclear Power Plants experimental data for the evaluation of the influence of the light water reactor (LWR) coolant environment and postulated chloride contaminations on crack development and fatigue have been determined in order to verify and extend the basis for a reliable estimation of the residual service life of reactor components. The aim of the research project was the investigation of the environmental effects at low strain rate conditions and the determination of the fatigue life under cyclic loading at uniaxial and multiaxial stress state. The quasi-static tensile tests (Constant Extension Rate Test, CERT) were performed using 3 low strain rates, each differing by about one order of magnitude (2.5.10 -3 , 3.1.10 -4 and 2.3.10 -5 %/s). The low cycle fatigue (LCF) experiments were conducted applying alternating tensile-compression loading with strain amplitudes of 0.3, 0.5 and 0.9 % at strain rates of 0.1 and 0.01 %/s (tests in air primarily 0.1 %/s). The cyclic notched tensile tests were carried out with a nominal axial strain in the notch root of 0.5 % at a strain rate of 0.1 %/s. The experiments in each case were performed in air, high purity water and chloride containing water at a testing temperature of 240 C, the oxygen content of the liquid medium was set to 0.4 ppm (simulated boiling water reactor coolant). In the CERT experiments chloride contents of 30, 50 and 100 ppb were applied, in the LCF tests the chloride content was 50 ppb which can be regarded as an upper realistic limit for a postulated chloride contamination of the reactor coolant. All experiments in liquid environment were preceded by a pre-autoclaving phase of at least 100 h in order to allow the formation of a stable oxide layer (magnetite). The testing material 22NiMoCr3-7 was available in form of an original reactor pressure vessel shell primarily designated for the German nuclear power plant

  10. Irradiated stainless steel material constitutive model for use in the performance evaluation of PWR pressure vessel internals

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.; Dunham, R.S. [ANATECH (United States); Demma, A. [Electric Power Research Institute - EPRI (United States)

    2011-07-01

    Demonstration of component functionality requires analytical simulations of reactor internals behavior. Towards that aim, EPRI has undertaken the development of irradiated material constitutive model and damage criteria for use in global and local finite-element based functionality analysis methodology. The constitutive behavioral regimes of irradiated stainless steel types 316 and 304 materials included in the model consist of: elastic-plastic material response considering irradiation hardening of the stress-strain curve, irradiation creep, stress relaxation, and void swelling. IASCC and degradation of ductility with irradiation are the primary damage mechanisms considered in the model. The material behavior model development consists of two parts: the first part is a user-material subroutine that can interface with a general-purpose finite element computer program to adapt it to the special-purpose of functionality analysis of reactor internals. The second part is a user utility in the form of Excel Spread sheets that permit users to extract a given property, e.g. the elastic-plastic stress-strain curve, creep curve, or void-swelling curve, as function of the relevant independent variables. The development of the model takes full advantage of the significant work that has been undertaken within EPRI's Material Reliability Program (MRP) to improve the knowledge of the material properties of irradiated stainless steels. Data from EPRI's MRP database have been utilized to develop equations that characterize the yield strength, ultimate tensile strength, uniform elongation, total elongation, reduction in area, void swelling and irradiation creep of stainless steels in a PWR environment. It is noted that, while the development of the model's equations has been statistically faithful to the material database, approximations were introduced in the model to ensure appropriate conservatism in the model's application consistently with accepted

  11. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

    2016-01-01

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  12. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Roychowdhury, S., E-mail: sroy27@gmail.com [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland); Materials Processing & Corrosion Engineering Division, Mod-Lab, D-Block, Bhabha Atomic Research Centre, Mumbai 400085 (India); Seifert, H.-P.; Spätig, P.; Que, Z. [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland)

    2016-09-15

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  13. Review of the models and mechanisms for environmentally-assisted crack growth of pressure vessel and piping steels in PWR environments

    International Nuclear Information System (INIS)

    Cullen, W.; Gabetta, G.; Hanninen, H.

    1985-12-01

    The crack-tip micromechanisms and the computational models for environmentally-assisted cracking in pressure vessel and piping steels in high-temperature, low-oxygen (PWR), reactor-grade water are described and evaluated in this report. The report begins with a brief description of the critical variables which are known to affect environmentally-assisted subcritical cracking in these metal/environment systems. The micromechanistic models are discussed in some detail, with anodic dissolution and hydrogen assistance being the prime candidates for the successful explanation of the observed phenomena. The anodic dissolution model offers far better quantification of the environmentally-assisted crack growth rates, but tends to overpredict the rates for a large number of conditions. The hydrogen assistance models qualitatively could account for a wider range of effects, but quantification of the model is virtually nonexistent. A variety of calculational models are in various stages of development; all of them are far from use as a predictive tool. Crack-tip strain rate models have received the most attention, and the approach to their use has been to partition the environmentally-assisted growth rates into a mechanically-driven component, with the environmental enhancement superposed. The environment component is then correlated with a calculated crack-tip strain rate. 141 refs., 59 figs

  14. Application of Response Surface Methodology for Modeling of Postweld Heat Treatment Process in a Pressure Vessel Steel ASTM A516 Grade 70.

    Science.gov (United States)

    Peasura, Prachya

    2015-01-01

    This research studied the application of the response surface methodology (RSM) and central composite design (CCD) experiment in mathematical model and optimizes postweld heat treatment (PWHT). The material of study is a pressure vessel steel ASTM A516 grade 70 that is used for gas metal arc welding. PWHT parameters examined in this study included PWHT temperatures and time. The resulting materials were examined using CCD experiment and the RSM to determine the resulting material tensile strength test, observed with optical microscopy and scanning electron microscopy. The experimental results show that using a full quadratic model with the proposed mathematical model is YTS = -285.521 + 15.706X1 + 2.514X2 - 0.004X1(2) - 0.001X2(2) - 0.029X1X2. Tensile strength parameters of PWHT were optimized PWHT time of 5.00 hr and PWHT temperature of 645.75°C. The results show that the PWHT time is the dominant mechanism used to modify the tensile strength compared to the PWHT temperatures. This phenomenon could be explained by the fact that pearlite can contribute to higher tensile strength. Pearlite has an intensity, which results in increased material tensile strength. The research described here can be used as material data on PWHT parameters for an ASTM A516 grade 70 weld.

  15. 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.

  16. J-R Fracture Resistance of SA533 Gr.B-Cl.1 Steel for Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji-Hyun; Hong, Seokmin; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A rolled plate might show different mechanical behaviors from a forging, even though they contain same chemical compositions. Furthermore, it is known that the fracture behavior of a rolled plate is very sensitive to material orientation comparing to a forging. In this study, the J-R fracture resistances of SA533 Gr.B-Cl.1 plate were measured at reactor operating temperature and the material orientation sensitivity was discussed. The decrease of fracture resistance of this kind of low alloy steel at an elevated temperature is known as the effect of dynamic strain aging (DSA). It was attributed to that the carbides and grains elongated to primary rolling direction, so that the aspect ratio of carbides and grains in the specimen with T-L orientation is larger. Generally, the hard second phase could take a roll of trigger point of unstable fracture. It is needed that the fracture surfaces of the tested specimens to be examined profoundly.

  17. The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

    International Nuclear Information System (INIS)

    Niffenegger, M.; Reichlin, K.; Kalkhof, D.

    2002-05-01

    The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

  18. 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)

  19. 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

  20. Fractographic observations of cleavage initiation in the ductile-brittle transition region of a reactor-pressure-vessel steel

    International Nuclear Information System (INIS)

    Rosenfield, A.R.; Shetty, D.K.; Skidmore, A.J.

    1983-01-01

    This note reports the results of a fractographic study conducted on a group of 1T compact fracture toughness specimens of a heavy-section A508 steel denoted TSE6 tested in the ductile-brittle transition region (22 and 82 0 C). The fatigue-precracked specimens were loaded at a rapid rate (760 or 550 mm per second) to promote cleavage-crack growth and lower-bound toughness behavior. All specimens experienced unstable cleavage fracture prior to reaching a maximum in the load displacement curve. Some ductile crack growth occurred in half of the specimens. The objective of fractographic examinations was to understand the observed statistical variations in cleavage initiation by (a) locating the origins of unstable cleavage fracture in the vicinity of the fatigue-precrack or ductilerupture crack fronts, (b) identifying microstructural features associated with the triggering of cleavage, and (c) documenting characteristic fracture surface dimensions such as the extent of stable-crack growth prior to unstable cleavage (Δα) and the distance of the cleavage origin from the ductilerupture front, /chi/ (or fatigue-crack front when Δα = 0)

  1. Application of the RKR model for evaluating the fracture toughness of pressure vessel steel in the transition temperature region

    International Nuclear Information System (INIS)

    Yang, Won Jon; Huh, Moo Young; Lee, Bong Sang; Hong, Jun Hwa

    2002-01-01

    Fracture toughness of a SA 533 B-1 steel was characterized in ductile-brittle transition temperature region by means of a RKR-type model. The original RKR model has been used to predict the plane strain fracture toughness (K IC ) behaviors in lower shelf region by assuming two material parameters, ie, the critical fracture stress and the characteristic distance. In this study, the fracture surface of every specimen was thoroughly investigated using scanning electron microscope to locate the actual cleavage initiation and to measure the cleavage initiation distance (CID) from the initial crack. The local fracture stress (σ f * ) of material was determined from the elastic-plastic stress field at the measured cleavage initiation location in the notched and precracked specimen. The local fracture stress of the precracked specimens was much higher than that of the notched specimen. The measured CIDs were strongly dependent on the test temperature and also on the fracture toughness. Based on the observations, it is found that, in the RKR-type cleavage fracture models, the characteristic distance should not be treated as a constant material parameter in the ductile-brittle transition region where the cleavage initiation controls the overall fracture process

  2. Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.K.

    2001-01-30

    A combined atom probe tomography and atom probe field ion microscopy study has been performed on a submerged arc weld irradiated to high fluence in the Heavy-Section Steel irradiation (HSSI) fifth irradiation series (Weld 73W). The composition of this weld is Fe - 0.27 at. % Cu, 1.58% Mn, 0.57% Ni, 0.34% MO, 0.27% Cr, 0.58% Si, 0.003% V, 0.45% C, 0.009% P, and 0.009% S. The material was examined after five conditions: after a typical stress relief treatment of 40 h at 607 C, after neutron irradiation to a fluence of 2 x 10{sup 23} n m{sup {minus}2} (E > 1 MeV), and after irradiation and isothermal anneals of 0.5, 1, and 168 h at 454 C. This report describes the matrix composition and the size, composition, and number density of the ultrafine copper-enriched precipitates that formed under neutron irradiation and the change in these parameters with post-irradiation annealing treatments.

  3. 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)

  4. 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

  5. Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

    International Nuclear Information System (INIS)

    Weiss, R.; Becker, R.; Lucht, B.; Mohr, F.; Hartwig, K.

    2001-01-01

    As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed. - Detection of surface connected longitudinal and transverse flaws: - notch, 3 mm deep and 10 mm long, for weave bead cladding; - notch, 2 mm deep and 20 mm long, for strip weld cladding. - Detection of embedded planar longitudinal and transverse flaws: - ligament of 7 mm for 8 mm clad thickness and 3 mm; - ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm. - Detection of embedded volumetric longitudinal and transverse flaws: - 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep

  6. 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.

  7. 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)

  8. 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.)

  9. 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

  10. 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

  11. 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

  12. 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

  13. Preliminary results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Matsumoto, T.; Komine, K.; Arai, S.

    1997-01-01

    A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11-12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented

  14. Heavy wall pressure vessels for energy systems

    International Nuclear Information System (INIS)

    Canonico, D.A.

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

  15. 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

  16. 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

  17. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  18. 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.

  19. Testing of a steel containment vessel model

    International Nuclear Information System (INIS)

    Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.; Komine, K.; Costello, J.F.

    1997-01-01

    A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

  20. Effect of Tempering Temperature on the Microstructure and Properties of Fe-2Cr-Mo-0.12C Pressure Vessel Steel

    Science.gov (United States)

    Wang, Qi-wen; Li, Chang-sheng; Peng, Huan; Chen, Jie; Zhang, Jian

    2018-03-01

    To obtain the high-temperature strength and toughness of the medium-high-temperature-pressure steel, the microstructure evolution and mechanical properties of Fe-2Cr-Mo-0.12C steel subjected to three different tempering temperatures after being normalized were investigated. The results show that the microstructure of the sample, tempered in the range 675-725 °C for 50 min, did not change dramatically, yet the martensite/austenite constituents decomposed, and the bainite lath merged together and transformed into polygonal ferrite. At the same time, the precipitate size increased with an increase in tempering temperature. With the increase in the tempering temperature from 675 to 725 °C, the impact absorbed energy of the Fe-2Cr-Mo-0.12C steel at -40 °C increased from 257 to 325 J, and the high-temperature yield strength decreased; however, the high-temperature ultimate tensile strength tempered at 700 °C was outstanding (422-571 MPa) at different tested temperatures. The variations of the properties were attributed to the decomposition of M/A constituents and the coarsening of the precipitates. Fe-2Cr-Mo-0.12C steel normalized at 930 °C and tempered at 700 °C was found to have the best combination of ductility and strength.

  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. 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. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2006-07-01

    Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

  4. Fabrication history and mechanical properties for ASTM A-533 Grade-B Class-2 steel weld for fully welded nuclear pressure vessels

    International Nuclear Information System (INIS)

    Pachur, D.

    1979-01-01

    Till now pressure vessels for light water reactors were made from rolled plates and forgings connected with each other by welding. The optimal quality of plates and forgings are limited in principle by the foundry technology. It is well known that in this process decomposition and segregation zones occur. Besides the heat affected zone created by the welding process is a weak link. The heat affected zone is heterogeneous and can be harbinger of risks leading to cracks. The production of a pressure vessel through shape welding is an alternative. The cylindrical container is produced by the application of one layer of welding after the other in a preshaped form. During the welding process the previously applied layers are simultaneously being tempered. The undesirable chemical residual elements are evenly distributed and segregation zones do not occur. Since we have only welding material the disadvantages of a heat affected zone are avoided. Furthermore the mechanical properties are independent of location and orientation. This shape welding process proved to be highly economical already during the experimental stay. Besides this process is applicable for vessel of any desired dimension

  5. 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.

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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.)

  14. 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

  15. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  16. 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

  17. The relevance of crack arrest phenomena for pressure vessel structural integrity assessment

    International Nuclear Information System (INIS)

    Connors, D.C.; Dowling, A.R.; Flewitt, P.E.J.

    1996-01-01

    The potential role of a crack arrest argument for the structural integrity assessments of steel pressure vessels and the relationship between crack initiation and crack arrest philosophies are described. A typical structural integrity assessment using crack initiation fracture mechanics is illustrated by means of a case study based on assessment of the steel pressure vessels for Magnox power stations. Evidence of the occurrence of crack arrest in structures is presented and reviewed, and the applications to pressure vessels which are subjected to similar conditions are considered. An outline is given of the material characterisation that would be required to undertake a crack arrest integrity assessment. It is concluded that crack arrest arguments could be significant in the structural integrity assessment of PWR reactor pressure vessels under thermal shock conditions, whereas for Magnox steel pressure vessels it would be limited in its potential to supporting existing arguments. (author)

  18. 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)

  19. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels; Etude a la sonde atomique de l`evolution microstructurale sous irradiation d`alliages ferritiques Fe-Cu et d`aciers de cuve REP

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends.

  20. 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.

  1. The fracture mechanical significance of cracks formed during stress-relief annealing of a submerged arc weldment in pressure vessel steel of type A508 class 2

    International Nuclear Information System (INIS)

    Liljestrand, L.-G.; Oestberg, G.

    1978-01-01

    In large weldments of type A508 C12 cracks can form in the heat-affected zone during stress-relief annealing. The significance of such cracks with respect to catastrophic fracture is of interest from the point of view of safety, in particular for nuclear pressure vessels. In this investigation the size of reheat cracks, as formed and after fatigue growth, has been compared with the critical size for fast fracture. The latter was assessed by determination of the toughness of the heat-affected zones. The fracture toughness of the heat-affected zones did not differ much from that of the parent material. The presence of microcracks reduced the fracture toughness (of a special type of simulated specimen) at 20 0 C by about 20%. The fracture mechanical evaluation indicates that the cracks formed during stress-relief annealing should not impair the safety of the vessel under normal conditions, except for particular geometries and when the cracks may rapidly link together during fatigue. (author)

  2. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1994-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

  3. 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.)

  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. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium-specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction

  6. R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

    2000-01-01

    Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study thye evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

  7. 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

  8. 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)

  9. 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.)

  10. 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)

  11. 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

  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. Experimental studies of oxidic molten corium-vessel steel interaction

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

    2001-01-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

  14. Experimental studies of oxidic molten corium-vessel steel interaction

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

    2001-12-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

  15. 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.

  16. Low temperature radiation embrittlement for reactor vessel steels

    International Nuclear Information System (INIS)

    Ginding, I.A.; Chirkina, L.A.

    1978-01-01

    General conceptions of cold brittleness of bcc metals are in a review. Considered are experimental data and theoretical representations about the effect of irradiation conditions, chemical composition, phase and structural constitutions, grain size, mechanical and thermomechanical treatments on low-temperature irradiation embrittlement of reactor vessel steels. Presented are the methods for increasing radiation stability of metals (carbon and Cr-Mo steels) used in manufacturing reactor vessels

  17. 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...

  18. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1993-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments

  19. Application of high strength steel to nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

    1976-01-01

    Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

  20. Study on Material Selection of Reactor Pressure Vessel of SCWR

    Science.gov (United States)

    Ma, Shuli; Luo, Ying; Yin, Qinwei; Li, Changxiang; Xie, Guofu

    This paper first analyzes the feasibility of SA-508 Grade 3 Class 1 Steel as an alternative material for Supercritical Water-Cooled Reactor (SCWR) Reactor Pressure Vessel (RPV). This kind of steel is limited to be applied in SCWR RPV due to its quenching property, though large forging could be accomplished by domestic manufacturers in forging aspect. Therefore, steels with higher strength and better quenching property are needed for SWCR RPV. The chemical component of SA-508 Gr.3 Cl.2 steel is similar to that of SA-508 Gr.3 Cl.1 steel, and more appropriate matching of strength and toughness could be achieved by the adjusting the elements contents, as well as proper control of tempering temperature and time. In light of the fact that Cl.2 steel has been successfully applied to steam generator, it could be an alternative material for SWCR RPV. SA-508 Gr.4N steel with high strength and good toughness is another alternative material for SCWR RPV. But large amount of research work before application is still needed for the lack of data on welding and irradiation etc.

  1. 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.

  2. 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.)

  3. 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

  4. 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.

  5. Development of improved SGV480 steel plate for containment vessel in PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Norioki [Advanced Nuclear Equipment Research Inst., Tokyo (Japan); Morikage, Yasushi; Okayama, Yutaka; Higashikubo, Tomohiro

    2001-01-01

    When a nuclear containment vessel made of steel plate at PWR plants in Japan is produced, SGV480 steel plate made by annealing method according to JIS G3118 is usually used in main. And, when thickness of welding portion of the vessel is larger than 38 mm, as heat treatment after welding is regulated to carry out according to the ministerial ordinance, it is difficult in actual to carry out the heat treatment of the actual welded portions. In a leading plant, approval of welding using a special method without heat treatment less than 47.25 mm of SGV480 carbon steel plate for JIS G3118 middle and ordinary pressure vessel was carried out to supply it for actual use. And, it is required for protection of welding fracture to carry out pre-heat treatment before welding. Because of increasing plate thickness requiring for lower temperature and more seismic resistance in construction condition, in order to produce a containment vessel without heat treatment after welding, more toughness is required for using material and welded portion. Therefore, a new SGV480 steel plate was developed by using TMCP method of modern steel manufacturing technology, to establish lower carbon equivalence and finer texture with upgrading of both toughness and weldability, without heat treatment after welding and pre-heat treatment before welding, at the Shin-Nippon Steel Co, Ltd. and Kawasaki Steel, Co. Ltd., respectively. (G.K.)

  6. 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.)

  7. 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.)

  8. Analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. 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

  10. 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

  11. 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.)

  12. 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

  13. 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

  14. 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.)

  15. 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.)

  16. Safety of light-water reactor pressure vessels against brittle fracture

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1979-01-01

    The results are surveyed of research by SKODA Trust into brittle failure resistance of materials for WWER type reactor pressure vessels and into pressure vessel operating safety. Conditions are discussed in detail decisive for initiation, propagation and arrest of brittle fracture. The tests on the Cr-Mo-V type steel showed high resistance of the steel to the formation and the propagation of brittle fracture. They also confirmed the high operating reliability and the required service life of the steel. (B.S.)

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

    Science.gov (United States)

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

    1973-01-01

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

  18. Comparison on Mechanical Properties of SA508 Gr.3 Cl.1, Cl.2, and Gr.4N Low Alloy Steels for Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Chul; Park, Sang-Gyu; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Ki-Hyoung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    In this study, microstructure and mechanical properties of SA508 Gr.3 Cl. 1, Cl.2, and Gr.4N low alloy steels are characterized to compare their properties. To evaluate the fracture toughness in the transition region, the master curve method according to ASTM E1921 was adopted in the cleavage transition region. Tensile tests and Charpy impact tests were also performed to evaluate the mechanical properties, and a microstructural investigation was carried out. The microstructure and mechanical properties of SA508 Gr.3 Cl.1, Cl2 and Gr.4N low alloy steels were characterized.. The predominant microstructure of SA508 Gr.4N model alloy is tempered martensite, while SA508 Gr.3 Cl.1 and Cl.2 steels show a typical tempered upper bainitic structure. SA508 Gr. 4N model alloy shows the best strength and transition behavior among the three SA508 steels. SA508 Gr.3 Cl.2 steel also has quite good strength, but there is a loss of toughness.

  19. 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

  20. 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

  1. 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

  2. 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.)

  3. 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)

  4. 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)

  5. 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)

  6. 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.)

  7. Monitoring of irradiation effects on the pressure vessel steels of Calder, Chapelcross and Windscale Advanced Gas Cooled Reactor (WAGR) nuclear reactors

    International Nuclear Information System (INIS)

    Turner, F.

    1980-01-01

    Tensile, Charpy and bend specimens of plate, forging and weld metal are exposed in the lower and upper zones of the reactors to neutron fluxes covering the range experienced by the vessels. The test conditions are described, the results presented and discussed. (author)

  8. 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.

  9. 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.

  10. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments

  11. Corrosion of vessel steel during its interaction with molten corium

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation)]. E-mail: bechta@sbor.spb.su; Khabensky, V.B. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Vitol, S.A. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Krushinov, E.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Granovsky, V.S. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Lopukh, D.B. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Gusarov, V.V. [Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), Odoevsky str., b. 24/2, 199155 St. Petersburg (Russian Federation); Martinov, A.P. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Martinov, V.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Fieg, G. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Tromm, W. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Bottomley, D. [Europaeische Kommission, General Direktion GFS, Institut fuer Transurane (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tuomisto, H. [Fortum Engineering Ltd. 00048 FORTUM, Rajatorpantie 8, Vantaa (Finland)

    2006-07-15

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments.

  12. 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

  13. 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

  14. Failure internal pressure of spherical steel containments

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, G.

    1985-01-01

    An application of the British CEGB's R6 Failure Assessment Approach to the determination of failure internal pressure of nuclear power plant spherical steel containments is presented. The presence of hypothetical cracks both in the base metal and in the welding material of the containment, with geometrical idealizations according to the ASME Boiler and Pressure Vessel Code (Section XI), was taken into account in order to analyze the sensitivity of the failure assessment with the values of the material fracture properties. Calculations of the elastoplastic collapse load have been performed by means of the Finite Element System SAMCEF. The clean axisymmetric shell (neglecting the influence of nozzles and minor irregularities) and two major penetrations (personnel and emergency locks) have been taken separately into account. Large-strain elastoplastic behaviour of the material was considered in the Code, using lower bounds of true stress-true strain relations obtained by testing a collection of tensile specimens. Assuming the presence of cracks in non-perturbed regions, the reserve factor for test pressure and the failure internal pressure have been determined as a function of the flaw depth. (orig.)

  15. 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

  16. Manufacturing of ultra-large diameter 20 MnMoNi 5 5 steel forgings for reactor pressure vessels and their properties

    International Nuclear Information System (INIS)

    Onodera, S.; Kawaguchi, S.; Tsukada, H.; Moritani, H.; Suzuki, K.; Sato, I.

    1985-01-01

    As the structural material for RPV typical of increased dimensions, as set of ultra-large diameter steel forgings for a PHWR RPV is presented as outlined below. (1) Material designation: 20 MnMoNi 5 5 (similar to SA508, Cl.3) (2) Size of the forgings: flanges, 8.440 mm OD, a weight of 238 tons for shell flange; shells and torus, 7,920 mm OD, with large height, cover dome, 6,800 mm OD in chord and 460 mm thick; blank before formed to dome is ca. 8,000 mm OD. (3) Chemical composition: particular effort was made for minimizing the tramp elements as P, S, As, Sn, Sb, Cu. (4) Manufacturing, key points: steel making - combined refining and degassing in ladle; ingot making - largest size ingots, including 570 ton and 500 ton ingots; forging - special ''outside-the-press'' forging and forming techniques; heat treatment - prevention of H 2 flaking in normalizing and tempering and handling of the extra-large forgings at water quenching. (5) Metallurgical properties: sufficiently uniform carbon distributions in the forgings; a lowest possible content of hydrogen, non-metallic inclusions and oxygen. Mechanical properties: uniformity in tensile and toughness properties; flaws - only limited number of spots of UT indications under 2 mm EFG (EFS). (orig.)

  17. Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2011-08-15

    Highlights: > Microstructural changes in stainless steel electroslag weld-overlay cladding. > Thermal aging caused progress of spinodal decomposition and precipitation of G phases in the {delta}-ferrite phase. > The degree of the spinodal decomposition had a linear relationship to the hardness. - Abstract: The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% {delta}-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 deg. C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M{sub 23}C{sub 6} type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

  18. 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.

  19. 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.

  20. Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

    1983-07-01

    The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

  1. Critical experiments, measurements and analyses to establish a crack arrest methodology for nuclear pressure vessel steels. Sixth quarterly progress report, January--March 1976

    International Nuclear Information System (INIS)

    Hahn, G.T.; Gehlen, P.C.; Hoagland, R.G.; Kanninen, M.F.; Popelar, C.; Rosenfield, A.R.

    1976-07-01

    Governing equations for a complete two-dimensional analysis of fast fracture and crack arrest (including inertia and thermal gradients) are derived. Strain energy and compliance values calculated for an SEN specimen agree with experiment. In the case of a nonstationary crack, the strain energy release rate is computed from the nodal displacements in the immediate vicinity of the crack tip. This value of G is sensitive to the spacings between nodes in the x and y direction. Results of two-dimensional analyses of propagation and arrest in the SEN test piece are also described; they are in general accord with one propagation event measured in an SEN A533B steel test piece. Compliance measurements were carried out on the rectangular DCB specimen to test the accuracy of the compliance values calculated with the one-dimensional beam-on-elastic-foundation model. These compliance values and the derivatives of compliance with crack length are used to relate the loads or displacements at fracture onset and at arrest to K/sub Q/, K/sub D/, and K/sub a/. Efforts were also made to test the assumption that energy losses arising from damping remote from the crack tip can be neglected. Measurements of damped oscillations in an A533B DCB specimen with a stationary crack indicate that total damping rates are relatively low, and that oscillations of the load pins in the pinhole account for the bulk of the damping. Finally, a new procedure for measuring K/sub D/ and K/sub m/ values is described; it is based on a dynamic analysis of the crack length at arrest and dynamically stiff wedge loading (the tie-down device) and eliminates the need for crack velocity measurements. It can also be applied to both small and large crack jumps and to ordinary and duplex-DCB specimens. Results for A533B steel agree with K/sub D/-values obtained from velocity measurements. The new procedure is also used to analyze results for a series of ship steels

  2. 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

  3. 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.

  4. 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.)

  5. Effect of microstructure on the impact toughness and temper embrittlement of SA508Gr.4N steel for advanced pressure vessel materials.

    Science.gov (United States)

    Yang, Zhiqiang; Liu, Zhengdong; He, Xikou; Qiao, Shibin; Xie, Changsheng

    2018-01-09

    The effect of microstructure on the impact toughness and the temper embrittlement of a SA508Gr.4N steel was investigated. Martensitic and bainitic structures formed in this material were examined via scanning electron microscopy, electron backscatter diffraction, transmission electron microscopy, and Auger electron spectroscopy (AES) analysis. The martensitic structure had a positive effect on both the strength and toughness. Compared with the bainitic structure, this structure consisted of smaller blocks and more high-angle grain boundaries (HAGBs). Changes in the ultimate tensile strength and toughness of the martensitic structure were attributed to an increase in the crack propagation path. This increase resulted from an increased number of HAGBs and refinement of the sub-structure (block). The AES results revealed that sulfur segregation is higher in the martensitic structure than in the bainitic structure. Therefore, the martensitic structure is more susceptible to temper embrittlement than the bainitic structure.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  7. 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)

  8. 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.)

  9. Fracture toughness of irradiated and recovered vessel steels

    International Nuclear Information System (INIS)

    Perosanz, F.; Lapena, J.

    1998-01-01

    This paper presents the fracture toughness measurements carried out on three vessel steels in an irradiated condition and after a post-irradiation recovery treatment. A statistical approach and the fracture parameters corresponding to two theoretical models of the fracture tests are used for evaluating toughness. Test results show that the neutron fluence gradually transforms the fracture behaviour of the vessel steels from ductile to brittle and seriously reduces their fracture toughness. The effectiveness of the recovery treatment, as evaluated from the toughness measurements, is confirmed, although the efficiency is not the same for the steels and depends on the evaluation parameter except in the case of almost complete recovery. The recovery effect increases with the received neutron fluence if the toughness values after treatment are compared with those in the irradiated condition rather than those in the as received condition. (orig.)

  10. 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.

  11. 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.

  12. Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Marini, B., E-mail: bernard.marini@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France); Averty, X. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SEMI (now DEN/DANS/DM2S/SEMT), F-91191 Gif-sur Yvette (France); Wident, P.; Forget, P.; Barcelo, F. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France)

    2015-10-15

    The hardening and the embrittlement under neutron irradiation of an A508 type RPV steel considering three different microstructures (bainite, bainite-martensite and martensite)have been investigated These microstructures were obtained by quenching after autenitization at 1100 °C. The irradiation induced hardening appears to depend on microstructure and is correlated to the yield stress before irradiation. The irradiation induced embrittlement shows a more complex dependence. Martensite bearing microstructures are more sensitive to non hardening embrittlement than pure bainite. This enhanced sensitivity is associated with the development of intergranular brittle facture after irradiation; the pure martensite being more affected than the bainite-martensite. It is of interest to note that this mixed microstructure appears to be more embrittled than the pure bainitic or martensitic phases in terms of temperature transition shift. This behaviour which could emerge from the synergy of the embrittlement mechanisms of the two phases needs further investigations. However, the role of microstructure on brittle intergranular fracture development appears to be qualitatively similar under neutron irradiation and thermal ageing.

  13. Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

    Science.gov (United States)

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

    2011-08-01

    The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% δ-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M 23C 6 type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

  14. A study on the fracture toughness of heavy section steel plates and forgings for nuclear pressure vessels produced in Japan, (4)

    International Nuclear Information System (INIS)

    Sakai, Yuzuru; Ogura, Nobukazu; Takahashi, Isao; Miya, Kenzo; Ando, Yoshio.

    1985-01-01

    As another parameter for evaluating the toughness of structural materials, there is crack arrest toughness. This is a parameter showing the resistance of materials to stop the cracks rapidly propagating in brittle state within the materials, unlike static and dynamic fracture toughness related to the occurrence of breaking. As the conventional method of determining the crack arrest toughness, the relatively large testing method such as double tensile test and ESSO test have been known, but the establishment of a smaller convenient testing method is desired. In this study, the evaluation of the crack arrest toughness of the very thick steel materials produced in Japan was carried out by the testing method using small test pieces. In order to make test pieces small, tapered type DCB test and the three-point bending test using DWTT test pieces were examined as well as the testing method recommended by ASTM. The test materials were A 533B, Cl. 1 and A 508, Cl. 3. The test pieces, the various testing methods and the experimental results are reported. The temperature dependence of the crack arrest toughness was shown. (Kako, I.)

  15. A study on the fracture toughness of heavy section steel plates and forgings for nuclear pressure vessels produced in Japan, 2

    International Nuclear Information System (INIS)

    Sakai, Yuzuru; Ogura, Nobukazu; Takahashi, Isao; Miya, Kenzo; Ando, Yoshio.

    1984-01-01

    In this paper, the main results of a series of tests carried out by the Atomic Energy Research Committee, the Japan Welding Engineering Society, for six years for the purpose of evaluating the fracture toughness and strength of superthick steel materials for nuclear reactors made in Japan are reported. In this research, as the fracture toughness test, three kinds of static, dynamic and crack propagation stop tests were carried out. Not only parent metals but also welded parts were evaluated, and numerous data have been accumulated. The fracture toughness of structural materials generally depends on test temperature, and forms three regions of lower shelf, transition and upper shelf from low temperature side toward high temperature side. It is desired to establish the effective method to determine fracture toughness over wide temperature range with small test pieces, and as its promising method, J(IC) fracture toughness test based on elasto-plastic fracture mechanics is carried out. The toughness in lower shelf and transition regions was clarified by K(IC) test, and that in upper shelf region was evaluated by J(IC) test. The methods of test and analysis, and the results are reported. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  17. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  18. 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...

  19. 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

  20. 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.)

  1. Mass optimization of a small pressure vessel using metal/FRP (fiber reinforced polymers) hybrid structures

    International Nuclear Information System (INIS)

    Nisar, J.A.; Abdullah, A.N.; Iqbal, N.

    2004-01-01

    In hybrid pressure vessels, composite (Fiber) is wound over a metallic liner (Steel/Aluminum) in hoop direction. In this concept of hybrid pressure vessel structure, metallic liner takes all the axial loads and fiber reinforced polymers (FRP/sub s/) takes load in circumferential (Hoop) direction. Hybrid structures combine the relatively high shear stiffness and ductility of metal alloy with high specific stiffness, strength and fatigue properties of FRP/sub s/. The relatively simple methods for producing hybrid structures circumvent the need for the complex and expensive equipment that is used for advanced composites processing. This paper presents an efficient way of designing a hybrid pressure vessel where prime concern is weight reduction over an equivalent aluminum structure and investigates various methodologies regarding combinations of metals and FRP/sub s/ for optimization of a given pressure vessel. For this purpose we adopted two different methods of simulation one is computer simulation using ANSYS and other is experimental verification by hydrostatic testing of manufactured pressure vessel. Two different pressure vessels one with aluminum liner and other with steel liner were fabricated. Kevlar 49/epoxy was wrapped around the liners in hoop direction. Both the pressure vessels were put into hydrostatic test. Strains were measured during the test and then converted into corresponding stresses. Results of hydrostatic test were quite in favor of the ANSYS results. In this way we have successfully designed, manufactured and tested the Hybrid pressure vessel saving almost 40% weight in case of aluminum liner and 43.6% in case of steel liner. (author)

  2. 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)

  3. 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

  4. Correlation between radiation damage and magnetic properties in reactor vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kempf, R.A., E-mail: kempf@cnea.gov.ar [División Caracterización, GCCN, CAC-CNEA (Argentina); Sacanell, J. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Milano, J. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Guerra Méndez, N. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Winkler, E.; Butera, A. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Troiani, H. [División Física de Metales, CAB-CNEA and Instituto Balseiro (UNCU), CONICET (Argentina); Saleta, M.E. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Fortis, A.M. [Departamento Estructura y Comportamiento. Gerencia Materiales-GAEN, CAC-CNEA (Argentina)

    2014-02-01

    Since reactor pressure vessel steels are ferromagnetic, provide a convenient means to monitor changes in the mechanical properties of the material upon irradiation with high energy particles, by measuring their magnetic properties. Here, we discuss the correlation between mechanical and magnetic properties and microstructure, by studying the flux effect on the nuclear pressure vessel steel used in reactors currently under construction in Argentina. Charpy-V notched specimens of this steel were irradiated in the RA1 experimental reactor at 275 °C with two lead factors (LFs), 93 and 183. The magnetic properties were studied by means of DC magnetometry and ferromagnetic resonance. The results show that the coercive field and magnetic anisotropy spatial distribution are sensitive to the LF and can be explained by taking into account the evolution of the microstructure with this parameter. The saturation magnetization shows a dominant dependence on the accumulated damage. Consequently, the mentioned techniques are suitable to estimate the degradation of the reactor vessel steel.

  5. Applicability of JIS SPV 50 steel to primary containment vessel of nuclear power station

    International Nuclear Information System (INIS)

    Iida, Kunihiro; Ishikawa, Koji; Sakai, Keiichi; Onozuka, Masakazu; Sato, Makoto.

    1979-01-01

    The space within reactor containment vessels must be expanded in order to improve the reliability of nuclear power plants, accordingly the adoption of large reactor containment vessels is investigated. SGV 42 and 49 steels in JIS G 3118 have been used for containment vessels so far, but stress relief annealing is required when the thickness exceeds 38 mm. The time has come when the use of thicker conventional plates without stress relieving or the use of high strength steel must be examined in detail. In this study, the tests of confirming material properties were carried out on SPV 50 in JIS G 3115, Steels for pressure vessels, aiming at the method of fabrication without stress relieving. The highest and lowest temperatures in use were set at 171 deg and -8 deg C, respectively. The chemical composition and the mechanical properties of the plates tested, the method of welding, the results of tensile test on the parent metal and the welds, the required lowest preheating temperature, the fracture toughness at low temperature and the brittle fracture causing test are reported. The parent metal and the welded joints of SPV 50 have the properties suitable to reactor containment vessels, namely the sufficient fracture toughness to guarantee the prevention of unstable fracture when the method of welding without stress relieving is adopted. (Kako, I.)

  6. 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.)

  7. Magnetic and electrical properties of ITER vacuum vessel steels

    International Nuclear Information System (INIS)

    Mergia, K.; Apostolopoulos, G.; Gjoka, M.; Niarchos, D.

    2007-01-01

    Full text of publication follows: Ferritic steel AISI 430 is a candidate material for the lTER vacuum vessel which will be used to limit the ripple in the toroidal magnetic field. The magnetic and electrical properties and their temperature dependence in the temperature range 300 - 900 K of AISI 430 ferritic stainless steels are presented. The temperature variation of the coercive field, remanence and saturation magnetization as well as electrical resistivity and the effect of annealing on these properties is discussed. (authors)

  8. 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.)

  9. Void nucleation by the helium atoms during lifetime of reactor pressure vessel

    International Nuclear Information System (INIS)

    Rahman, F.A.

    1984-01-01

    Void formation and growth has a great influence on the reactor pressure vessel steels during its lifetime and during post-irradiation annealing to increase its life. The present investigation aimed at the fact that if one can prevent void nucleation, accordingly one would not wary about void formation and growth. From that concept a model for helium production by transmutation reaction and corresponding swelling under irradiation conditions for several number of steels have been developed. This was done for recommending a steel type that can oppose such a phenomena. In the same time the present investigation gives a procedure utilizing such phenomena for checking the validity of pressure vessel steel used in the NPP

  10. 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

  11. 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)

  12. 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.

  13. Critical experiments, measurements and analyses to establish a crack arrest methodology for nuclear pressure vessel steels. Task 62. Second annual progress report, 1 July 1975 to 30 June 1976

    International Nuclear Information System (INIS)

    Hahn, G.T.; Gehlen, P.C.; Hoagland, R.G.; Marschall, C.W.; Kanninen, M.F.; Popelar, C.; Rosenfield, A.R.

    1976-10-01

    Developments of the theory, the analyses and the materials characterization for assessments of crack arrest in heavy walled nuclear pressure vessels are described. A two-dimensional, dynamic, finite difference analysis of crack propagation and arrest, including inertia forces and thermal stresses is derived. The analysis is used to treat run-arrest events in DCB- and SEN-specimens. Calculations based on the one-dimensional analysis are compared with detailed measurements of run-arrest in two polymeric materials. The agreements between calculations and experiments support the dynamic, energy conservation theory of crack arrest. Implications of a K/sub ID/-crack velocity curve with a negative slope are examined. The quantity K/sub Im/, the minimum propagating crack toughness, is identified as the crack arrest material property for engineering applications. Considerations entering the design of a crack arrest property measuring procedure, including specimen shape, size, thickness and the crack jump are discussed. The influence of side grooves on the propensity for crack branching is examined. Compliance measurements for rectangular DCB-specimens are presented and analyzed. Energy losses arising from damping remote from the crack tip in an A533B steel DCB specimen are described, and classed as negligible. Interactions between test pieces and the loading system are examined. A simplified procedure for measuring K/sub D/-, K/sub m/-, and K/sub a/-values, based on the dynamic analysis of the crack length at arrest is demonstrated. Measurements of K/sub D/ at temperatures in the range (NDT-66)C to (NDT +34)C and for crack velocities in the range 200 ms -1 to 1000 ms -1 are presented. The connection between propagating crack toughness values and the rapidly loaded stationary crack toughness K/sub Id/ is examined

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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)

  1. 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.

  2. 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

  3. 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)

  4. Corrosion fatigue crack growth of pressure vessel welds in PWR environment

    International Nuclear Information System (INIS)

    Bamford, W.H.; Ceschini, L.J.; Moon, D.M.

    1983-01-01

    The fatigue crack growth rate behavior of several pressure vessel steel welds in PWR environment is discussed. The behavior is compared with associated heat-affected zone behavior, and with comparable base metal results. The welds show different degrees of susceptibility to the environmental influence, and this is discussed in some detail, along with fractographic observations on the tested specimens

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  6. 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)

  7. 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.)

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  13. 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)

  14. 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

  15. 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.)

  16. 46 CFR 42.09-30 - Additional survey requirements for steel-hull vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Additional survey requirements for steel-hull vessels...-30 Additional survey requirements for steel-hull vessels. (a) In addition to the requirements in § 42...) When the vessel is in drydock, the hull plating, etc., shall be examined. (c) The holds, 'tween decks...

  17. A fracture mechanics approach to predicting the effects of warm prestressing and its applications to pressure vessels

    International Nuclear Information System (INIS)

    Chell, G.G.

    1979-01-01

    A theory of warm prestressing based on the J-integral is described. The theory is validated using experimental warm pre-stressing data obtained on a carbon-manganese steel, two pressure vessel steels and mild steel. The theory is applied to the pressurised water reactor and the effects of warm prestressing evaluated after irradiation damage to the pressure vessel, and in the case of a loss of coolant accident. Warm prestressing increases the resistance to inhibits the initiation and propagation of the cracks. The benefits of warm prestressing for shallow cracks is less certain and a more detailed analysis is required. (orig.)

  18. 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

  19. 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

  20. DOMPAC dosimetry experiment. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damages

    International Nuclear Information System (INIS)

    Alberman, A.; Faure, M.; Thierry, M.; Hoclet, O.; Le Dieu de Ville, A.; Nimal, J.C.; Soulat, P.

    1979-01-01

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI), was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel [fr

  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. 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

  3. 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

  4. Filament wound pressure vessels with load sharing liners for space shuttle orbiter applications

    International Nuclear Information System (INIS)

    Ecord, G.M.

    1976-01-01

    Early in the development of orbiter propulsion and environmental control subsystems it was recognized that use of overwrapped pressure vessels with load sharing liners may provide significant weight savings for high pressure gas containment. A program is described which was undertaken by Rockwell International to assess the utility for orbiter applications of titanium 6Al--4V and Inconel 718 liners overwrapped with Kevlar fibers. Also briefly described are programs administered by the NASA Lewis Research Center to evaluate cryoformed steel liners overwrapped with Kevlar fibers and to establish a method that can guarantee cyclic life of the vessels

  5. Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

  6. 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)

  7. 46 CFR 176.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

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

  8. 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.)

  9. 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

  10. 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)

  11. Concept of a Prestressed Cast Iron Pressure Vessel for a Modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Steinwarz, Wolfgang; Bounin, Dieter

    2014-01-01

    High Temperature Reactors (HTR) are representing one of the most interesting solutions for the upcoming generation of nuclear technology, especially with view to their inherent safety characteristics. To complete the safety concept of such plants already in the first phase of the technical development, Prestressed Cast Iron Pressure Vessels (PCIV) instead of the established forged steel reactor pressure vessels have been considered under the aspect of safety against bursting. A longterm research and development work, mainly performed in Germany, showed the excellent features of this technical solution. Diverse prototypic vessels were tested and officially proven. Design studies confirmed the feasibility of such a vessel concept also for Light Water Reactor types, too. The main concept elements of such a burst-proof vessel are: Strength and tightness functions are structurally separated. The tensile forces are carried by the prestressing systems consisting of a large number of independent wires. Compressive forces are applied to the vessel walls and heads. These are segmented into blocks of ductile cast iron. All cast iron blocks are prestressed to high levels of compression. The sealing function is assigned to a steel liner fixed to the cast iron blocks. The prestressing system is designed for an ultimate pressure of 2.3 times the design pressure. The prestress of the lids is designed for gapping at a much smaller pressure. Therefore, a drop of pressure will always occur before loss of strength (“leakage before failure”). In addition to these safety features further technical as well as economic aspects generate favorable assessment criteria: high design flexibility, feasibility of large vessel diameters; advantageous conditions for transport, assembly and decommissioning due to the segmented construction; advantage of workshop manufacturing; high-level quality control of components. Nowadays, considering the globally newly standardized safety requirements

  12. 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

  13. Flaw density examinations of a clad boiling water reactor pressure vessel segment

    International Nuclear Information System (INIS)

    Cook, K.V.; McClung, R.W.

    1986-01-01

    Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel [nominally 0.7 by 3 m (2 by 10 ft)]. This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected

  14. High pressure deuterium-tritium gas target vessels for muon-catalyzed fusion experiments

    International Nuclear Information System (INIS)

    Caffrey, A.J.; Spaletta, H.W.; Ware, A.G.; Zabriskie, J.M.; Hardwick, D.A.; Maltrud, H.R.; Paciotti, M.A.

    1989-01-01

    In experimental studies of muon-catalyzed fusion, the density of the hydrogen gas mixture is an important parameter. Catalysis of up to 150 fusions per muon has been observed in deuterium-tritium gas mixtures at liquid hydrogen density; at room temperature, such densities require a target gas pressure of the order of 1000 atmospheres (100 MPa, 15,000 psi). We report here the design considerations for hydrogen gas target vessels for muon-catalyzed fusion experiments that operate at 1000 and 10,000 atmospheres. The 1000 atmosphere high pressure target vessels are fabricated of Type A-286 stainless steel and lined with oxygen-free, high-conductivity (OFHC) copper to provide a barrier to hydrogen permeation of the stainless steel. The 10,000 atmosphere ultrahigh pressure target vessels are made from 18Ni (200 grade) maraging steel and are lined with OFHC copper, again to prevent hydrogen permeation of the steel. In addition to target design features, operating requirements, fabrication procedures, and secondary containment are discussed. 13 refs., 3 figs., 1 tab

  15. 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.

  16. 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.)

  17. A determination of the benefits of annealing irradiated pressure vessel weldments

    International Nuclear Information System (INIS)

    Lott, R.G.; Mager, T.R.

    1988-01-01

    The long-term benefit of annealing an irradiated reactor pressure vessel steel may be described in terms of a benefit factor, B. The benefit factor compares the mechanical properties of an annealed and reirradiated specimen with an equivalent specimen having no intermediate anneal. The benefit factor was determined using a series of microhardness specimens prepared from nuclear pressure vessel surveillance program materials. These specimens were annealed and then reirradiated in a test reactor. There was an obvious long-term benefit in the specimens annealed at 450 0 C. The long-term benefit was less obvious at 400 0 C and no significant benefit was noted at 350 0 C. The benefit factor may also be used as the basis of a surveillance program for an annealed pressure vessel. A strategy for such a surveillance program is described. (author)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  19. 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

  20. Review of in-service thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1984-01-01

    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. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (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 (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)