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Sample records for vessel rpv steels

  1. Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels

    Energy Technology Data Exchange (ETDEWEB)

    Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

    2013-09-01

    The principle work at the atomic scale is to develop a predictive quantitative model for the microstructure evolution of RPV steels under thermal aging and neutron radiation. We have developed an AKMC method for the precipitation kinetics in bcc-Fe, with Cu, Ni, Mn and Si being the alloying elements. In addition, we used MD simulations to provide input parameters (if not available in literature). MMC simulations were also carried out to explore the possible segregation/precipitation morphologies at the lattice defects. First we briefly describe each of the simulation algorithms, then will present our results.

  2. Reactor Pressure Vessel (RPV) Acquisition Strategy

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    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

  3. Initial evaluation of ultrasonic attenuation measurements for estimating fracture toughness of RPV steels

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    Hiser, A.L. Jr.; Green, R.E. Jr. [Johns Hopkins Univ., Baltimore, MD (United States). Center for Nondestructive Evaluation

    1999-08-01

    Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently, there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides initial results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. (orig.)

  4. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels; RPV-1: un premier reacteur virtuel pour simuler les effets d'irradiation dans les aciers de cuve des reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Jumel, St

    2005-01-15

    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.)

  5. Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

    Energy Technology Data Exchange (ETDEWEB)

    Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

    2015-12-15

    Highlights: • German RPV steels were originally studied by positron annihilation spectroscopy. • Neutron irradiated and hydrogen ion implanted specimens were studied. • Both irradiation ways caused to increase of defect size. • We determined that the defect size was higher in implanted specimens. - Abstract: Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

  6. Effect of lead factors on the embrittlement of RPV SA-508 cl 3 steel

    Energy Technology Data Exchange (ETDEWEB)

    Kempf, Rodolfo, E-mail: kempf@cnea.gov.ar [CNEA, Unidad Actividad Combustibles Nucleares, División Caracterización, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina); Troiani, Horacio, E-mail: troiani@cab.cnea.gov.ar [Centro Atómico Bariloche (CNEA) e Instituto Balseiro (UNCU), CONICET, Av. Bustillo 9500, CP 8400, Rio Negro (Argentina); Fortis, Ana Maria, E-mail: fortis@cnea.gov.ar [CNEA, Departamento Estructura y Comportamiento, UNSAM, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina)

    2013-03-15

    This paper presents a project to study the effect of lead factors on the mechanical behaviour of the SA-508 type 3 Reactor Pressure Vessel (RPV) steel used in the reactor under construction Atucha II in Argentina. Charpy-V notch specimens of this steel were irradiated at the RA1 experimental reactor at a temperature of 275 °C with two lead factors (186 and 93). The neutron flux was 3.71 × 10{sup 15} n m{sup −2} s{sup −1} and 1.85 × 10{sup 15} n m{sup −2} s{sup −1} (E > 1 MeV) respectively. In both cases, the fluence was 6.6 × 10{sup 21} n m{sup −2}, which is equivalent to that received by the PHWR Atucha II RPV in 10 years of full power irradiation. The results of Charpy tests revealed significant embrittlement both in the ΔT = 14 °C and ΔT = 21 °C shifts of the ductile–brittle transition temperatures (DBTT) and in the reduction of the maximum energy absorbed. This result shows that the shift of the DBTT with a lead factor of 93 is larger than that obtained with a lead factor of 186. Then, the results of irradiation in experimental reactors (MTR) with high lead factors may not be conservative with respect to the actual RPV embrittlement.

  7. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

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    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.

  8. Comparative study of irradiated and hydrogen implantation damaged German RPV steels from PAS point of view

    Energy Technology Data Exchange (ETDEWEB)

    Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

    2014-09-01

    Highlights: • We used positron annihilation spectroscopy to study German RPV steels. • We examine microstructural changes in the studied specimens. • Simulation of neutron irradiation was performed by hydrogen ion implantation. • No large voids or vacancy clusters were formed due to irradiation or implantation. - Abstract: Commercial German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was also in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1–2 vacancies with relatively small contribution (with intensity on the level of 20–40%) were observed in “as-received” steels. A significant increase in the size of the induced defects due to neutron damage was observed at a level of 2–3 vacancies in the irradiated specimens. The size and intensity of defects reached a similar level as in the specimens irradiated in nuclear reactor due to hydrogen ions implantation with energy of 100 keV (up to the depth <500 nm). This could confirm the ability to simulate neutron damage by ion implantation.

  9. Comparative study of irradiated and hydrogen implantation damaged German RPV steels from PAS point of view

    Science.gov (United States)

    Pecko, Stanislav; Sojak, Stanislav; Slugeň, Vladimír

    2014-09-01

    Commercial German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was also in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40%) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed at a level of 2-3 vacancies in the irradiated specimens. The size and intensity of defects reached a similar level as in the specimens irradiated in nuclear reactor due to hydrogen ions implantation with energy of 100 keV (up to the depth <500 nm). This could confirm the ability to simulate neutron damage by ion implantation.

  10. Review and Assessment of SCC Experiments with RPV Steels in Oskarshamn 2 and 3 (ABB Report SBR 99-020)

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, Hans-Peter; Ritter, Stefan [Paul Scherrer Inst., Laboratory for Materials Behaviour, Villigen (Switzerland). Nuclear Energy and Safety Research Dept.

    2005-11-15

    Some years ago SKI and the Swedish utilities sponsored SCC investigations, where non cladded and cladded (Inconel 182/AISI 308L) bolt-loaded C(T) specimens of different reactor pressure vessel (RPV) steels have been exposed to boiling water reactor (BWR)/normal water chemistry (NWC) and hydrogen water chemistry (HWC) environment in Oskarshamn 3 and 2 during a five- and four-year period. In the following report, the Swedish stress corrosion cracking (SCC) data from this project are critically reviewed and assessed on the basis of the relevant service experience and of the accumulated experimental background knowledge on SCC of carbon (C) and low-alloy steel (LAS) and dissimilar weld joints in high-temperature (HT) water. The investigations in Oskarshamn 3/2 generally revealed a low SCC crack growth susceptibility of RPV steels and interdendritic (ID) SCC in the Inconel 182 weld metal under BWR/NWC and HWC conditions. All non-cladded specimens and specimens with AISI 308L stainless steel cladding revealed no or only minor crack growth. However, the specimens with Inconel 182 cladding and fatigue pre-crack in the weld metal revealed clear, but minor crack growth into the heat-affected zone (HAZ) of the RPV steel under BWR/NWC conditions. Marked crack growth (of 2.43 mm) in the RPV steel was only observed in one of the Inconel 182 cladded specimens tested under BWR/NWC conditions with a high K{sub I} value of 48.8 MPa.{radical}m, where the fatigue pre-crack-tip was located in the RPV steel base metal far beyond its HAZ. Although the extent of cracking has rather surprising for a bolt-loaded specimen, the average SCC crack growth rate (CGR) of 0.5 mm/year over the five-year testing period was still within the upper range of constant load SCC CGRs in autoclave tests in oxygenated high-purity water and below the BWRVIP-60 SCC disposition line (DL)1, and thus does not represent an immediate concern. The initial K{sub I} value of this specimen represents a rather deep crack

  11. Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

    Science.gov (United States)

    Pecko, Stanislav; Sojak, Stanislav; Slugeň, Vladimír

    2015-12-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

  12. Mechanical properties and microstructure of long term thermal aged WWER 440 RPV steel

    Science.gov (United States)

    Kolluri, M.; Kryukov, A.; Magielsen, A. J.; Hähner, P.; Petrosyan, V.; Sevikyan, G.; Szaraz, Z.

    2017-04-01

    The integrity assessment of the Reactor Pressure Vessel (RPV) is essential for the safe and Long Term Operation (LTO) of a Nuclear Power Plant (NPP). Hardening and embrittlement of RPV caused by neutron irradiation and thermal ageing are main reasons for mechanical properties degradation during the operation of an NPP. The thermal ageing-induced degradation of RPV steels becomes more significant with extended operational lives of NPPs. Consequently, the evaluation of thermal ageing effects is important for the structural integrity assessments required for the lifetime extension of NPPs. As a part of NRG's research programme on Structural Materials for safe-LTO of Light Water Reactor (LWR) RPVs, WWER-440 surveillance specimens, which have been thermal aged for 27 years (∼200,000 h) at 290 °C in a surveillance channel of Armenian-NPP, are investigated. Results from the mechanical and microstructural examination of these thermal aged specimens are presented in this article. The results indicate the absence of significant long term thermal ageing effect of 15Cr2MoV-A steel. No age hardening was detected in aged tensile specimens compared with the as-received condition. There is no difference between the impact properties of as-received and thermal aged weld metals. The upper shelf energy of the aged steel remains the same as for the as-received material at a rather high level of about 120 J. The T41 value did not change and was found to be about 10 °C. The microstructure of thermal aged weld, consisting carbides, carbonitrides and manganese-silicon inclusions, did not change significantly compared to as-received state. Grain-boundary segregation of phosphorus in long term aged weld is not significant either which has been confirmed by the absence of intergranular fracture increase in the weld. Negligible hardening and embrittlement observed after such long term thermal ageing is attributed to the optimum chemical composition of 15Cr2MoV-A for high thermal stability.

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

    Science.gov (United States)

    Wang, Xuejiao; Qiang, Wenjiang; Shu, Guogang

    2017-08-01

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

  14. Effects of the Microstructure on Segregation behavior of Ni-Cr-Mo High Strength Low Alloy RPV Steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    SA508 Gr.4N Ni-Cr-Mo low alloy steel has an improved fracture toughness and strength, compared to commercial Mn-Mo-Ni low alloy RPV steel SA508 Gr.3. Higher strength and fracture toughness of low alloy steels could be achieved by adding Ni and Cr. So there are several researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature and time of a reactor pressure vessel is more than 300 .deg. C and over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, it requires a resistance of thermal embrittlement in the high temperature range including temper embrittlement resistance. S. Raoul reported that the susceptibility to temper embrittlement was increasing a function of the cooling rate in SA533 steel, which suggests the martensitic microstructures resulting from increased cooling rates are more susceptible to temper embrittlement. However, this result has not been proved yet. So the comparison of temper embrittlement behavior was made between martensitic microstructure and bainitic microstructure with a viewpoint of boundary features in SA508 Gr.4N, which have mixture of tempered bainite/martensite. In this study, we have compared temper embrittlement behaviors of SA508 Gr.4N low alloy steel with changing volume fraction of martensite. The mechanical properties of these low alloy steels) were evaluated after a long-term heat treatment(450 .deg. C, 2000hr. Then, the images of the segregated boundaries were observed and segregation behavior was analyzed by AES. In order to compare the misorientation distributions of model alloys, grain boundary structures were measured with EBSD

  15. 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.

  16. In-situ magnetic property measurement of the RPV steel under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ebine, N.; Suzuki, M. [Japan Atomic Research Energy Inst. (Japan); Kikuchi, H.; Kamada, Y.; Ara, K.; Takahashi, S. [NDE and S-RC, Faculty of Engineering, Iwate Univ. (Japan)

    2004-07-01

    Magnetic properties of a RPV steel were successfully measured in-situ under neutron irradiation. Coercivity and Hysteresis loss increased with neutron fluence and showed a peak at the fluence of about 3 x 10{sup 19} n/cm{sup 2}, and the maximum increase of coercivity was 1.25%. The reason of showing a peak is not known. (orig.)

  17. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

    2017-07-01

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  18. Weld residual stresses near the bimetallic interface in clad RPV steel: A comparison between deep-hole drilling and neutron diffraction data

    Energy Technology Data Exchange (ETDEWEB)

    James, M.N., E-mail: mjames@plymouth.ac.uk [School of Marine Science and Engineering, University of Plymouth, Drake Circus, Plymouth (United Kingdom); Department of Mechanical Engineering, Nelson Mandela Metropolitan University, Port Elizabeth (South Africa); Newby, M.; Doubell, P. [Eskom Holdings SOC Ltd, Lower Germiston Road, Rosherville, Johannesburg (South Africa); Hattingh, D.G. [Department of Mechanical Engineering, Nelson Mandela Metropolitan University, Port Elizabeth (South Africa); Serasli, K.; Smith, D.J. [Department of Mechanical Engineering, University of Bristol, Queen' s Building, University Walk, Bristol (United Kingdom)

    2014-07-01

    Highlights: • Identification of residual stress trends across bimetallic interface in stainless clad RPV. • Comparison between deep hole drilling (DHD – stress components in two directions) and neutron diffraction (ND – stress components in three directions). • Results indicate that both techniques can assess the trends in residual stress across the interface. • Neutron diffraction gives more detailed information on transient residual stress peaks. - Abstract: The inner surface of ferritic steel reactor pressure vessels (RPV) is clad with strip welded austenitic stainless steel primarily to increase the long-term corrosion resistance of the ferritic vessel. The strip welding process used in the cladding operation induces significant residual stresses in the clad layer and in the RPV steel substrate, arising both from the thermal cycle and from the very different thermal and mechanical properties of the austenitic clad layer and the ferritic RPV steel. This work measures residual stresses using the deep hole drilling (DHD) and neutron diffraction (ND) techniques and compares residual stress data obtained by the two methods in a stainless clad coupon of A533B Class 2 steel. The results give confidence that both techniques are capable of assessing the trends in residual stresses, and their magnitudes. Significant differences are that the ND data shows greater values of the tensile stress peaks (∼100 MPa) than the DHD data but has a higher systematic error associated with it. The stress peaks are sharper with the ND technique and also differ in spatial position by around 1 mm compared with the DHD technique.

  19. Effects of the phase fractions on the carbide morphologies, Charpy and tensile properties in SA508 Gr.4N High Strength Low Alloy RPV Steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    To improve the strength and toughness of RPV (reactor pressure vessel) steels for nuclear power plants, an effective way is the change of material specification from tempered bainitic SA508 Gr.3 Mn-Mo-Ni low alloy steel into tempered martensitic/bainitic SA508 Gr.4N Ni-Cr-Mo low alloy steel. It is known that the phase fractions of martensitic/bainitic steels are very sensitive to the austenitizing cooling rates. Kim reported that there are large differences of austenitizing cooling rates between the surface and the center locations in RPV due to its thickness of 250mm. Hence, the martensite/bainite fractions would be changed in different locations, and it would affect the microstructure and mechanical properties in Ni-Cr-Mo low alloy steel. These results may lead to inhomogeneous characteristics after austenitizing. Therefore, it is necessary to evaluate the changes of microstructure and mechanical properties with varying phase fractions in Ni-Cr-Mo low alloy steel. In this study, the effects of martensite/bainite fractions on microstructure and mechanical properties in Ni-Cr-Mo low alloy steel were examined. The changes in phase fractions of Ni-Cr-Mo low alloy steel with different cooling rates were analyzed, and then the phase fractions were correlated with its microstructural observation and mechanical properties

  20. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  1. Comparison of the segregation behavior between tempered martensite and tempered bainite in Ni-Cr-Mo high strength low alloy RPV steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Kim, Min Chul; Kim, Hyung Jun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    SA508 Gr.4N Ni-Cr-Mo low alloy steel has an superior fracture toughness and strength, compared to commercial Mn-Mo-Ni low alloy RPV steel SA508 Gr.3. Higher strength and fracture toughness of low alloy steels could be obtained by adding Ni and Cr. So several were performed on researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature and term of a reactor pressure vessel is more than 300 .deg. C and over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, the resistance of thermal embrittlement in the high temperature range including temper embrittlement is required. S. Raoul reported that the susceptibility to temper embrittlement was increasing a function of the cooling rate in SA533 steel, which suggests the martensitic microstructures resulting from increased cooling rates are more susceptible to temper embrittlement. However, this result has not been proved yet. So the comparison of temper embrittlement behavior was made between martensitic microstructure and bainitic microstructure with a viewpoint of boundary features in SA508 Gr.4N, which have mixture of tempered bainite/martensite. We have compared temper embrittlement behaviors of SA508 Gr.4N low alloy steel with changing volume fraction of martensite. The mechanical properties of these low alloy steels were evaluated after a long-term heat treatment. Then, the the segregated boundaries were observed and segregation behavior was analyzed by AES. In order to compare the misorientation distributions of model alloys, grain boundary structures were measured with EBSD

  2. 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

  3. Study on the microstructure and toughness of RPV SA508 class 3 steel weldments

    Energy Technology Data Exchange (ETDEWEB)

    Ko, J. H. [Korea Univesity of Technology and Education, Cheonan (Korea, Republic of); Gang, Y. H.; Joo, K. N.; Hwang, Y. H. [KAERI, Taejon (Korea, Republic of); Kim, J. T.; Kwon, H. K. [Doosan Heavy Industries and Construction Company, Chanwon (Korea, Republic of)

    2003-10-01

    The microstructures of RPV SA508 class 3 steel multipass weld metals with submerged arc welding process by varying the heat inputs of 2.4 and 3.6kJ/mm were investigated by optical and scanning electron microscopes. The microstructures were also compared between as-welded and postweld heat treatment conditions. The relationship between weld microstructures and toughness as well as hardness of weld metals was evaluated. The toughness was enhanced a little in the lower heat input of 2.4kJ/mm but the hardness of welds was decreased. The microstructures of welds made at the lower heat input used in this study consisted of a little higher proportion of acicular ferrite than those of welds made at the higher heat input(3.6kJ/mm), in which unfavorable microstructure to toughness such as grain boundary ferrite and banitic structure were increased.

  4. Characterization by notched and precracked Charpy tests of the in-service degradation of RPV steel fracture toughness

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.

    1997-01-01

    The current engineering and regulatory practice to estimate fracture toughness safety margins for nuclear reactor pressure vessels (RPVs) relies heavily on the CVN impact test. Techniques to estimate in-service toughness degradation directly using a variety of precracked specimens are under development worldwide. Emphasis is on their miniaturization. In the nuclear context, it is essential to address many issues such as representativity of the surveillance programs with respect to the vessel in terms of materials and environment, transferability of test results to the structure (constraint and size effects), lower bound toughness certification, creadibility relative to trends of exising databases. An enhanced RPV surveillance strategy in under development in Belgium. It combines state-of-the-art micromechanical and damage modelling to the evaluation of CVN load-deflection signals, tensile stress-strain curves and slow-bend tests of reconstituted precracked Charpy specimens. A probabilistic micromechanical model has been established for static and dynamic transgranular cleavage initiation fracture toughness in the ductile-brittle transition temperature range. This model allows to project toughness bounds for any steel embrittlement condition from the corresponding CVN and static tensile properties, using a single scaling factor defined by imposing agreement with toughness tests in a single condition. The outstanding finding incorporated by this toughness transfer model is that the microcleavage fracture stress is affected by temperature in the ductile-brittle transition and that this influence is strongly correlated to the flow stress: this explains the shape of the K{sub Ic}n K{sub Id} temperature curves as well as the actual magnitude of the strain rate and irradiation effects. Furthermore, CVN crack arrest loads and fracture appearance are also taken advantage of in order to estimate K{sub Ia} degradation. Finally, the CVN-tensile load-temperature diagram

  5. Comparison of the Microstructure and Segregation behavior between SA508 Gr.3 and SA508 Gr.4N High Strength Low Alloy RPV Steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    It is generally known that SA508 Gr.4N low alloy steel has an improved fracture toughness and strength, compared to commercial low alloy steels such as SA508 Gr.3 which have lower than 1% Ni. Higher strength and fracture toughness of low alloy steels could be achieved by adding Ni and Cr. So there are several researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature and time of a reactor pressure vessel is more than 300 .deg. C and over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, it requires a phase stability in the high temperature range including temper embrittlement resistance. S. Raoul reported that the susceptibility to temper embrittlement was increasing an function of the cooling rate in SA533 steel, which suggests the martensitic microstructures resulting from increased cooling rates are more susceptible to temper embrittlement. However, this result has not been proved yet. So comparison was made between the temper embrittlement behaviors of SA508 Gr.3 and Gr.4N low alloy steel with a viewpoint of boundary features, which have different microstructures of tempered bainite(SA508 Gr.3) and tempered martensite(SA508 Gr.4N). In this study, we have compared temper embrittlement behaviors of SA508 Gr.3 and SA508 Gr.4N low alloy steel. The mechanical properties of these low alloy steels after a long-term heat treatment(450 .deg. C, 2000hr) were evaluated. Then, the images of the fracture surfaces were observed and grain boundary segregation was analyzed by AES. In order to compare the misorientation distributions of two model alloys, the grain boundary structures of the low alloy steels with EBSD were measured

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

    Energy Technology Data Exchange (ETDEWEB)

    McHenry, H.I.; Alers, G.A. [National Inst. of Standards and Technology, Boulder, CO (United States). Materials Reliability Div.

    1998-03-01

    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. Application of Bimodal Master Curve Approach on KSNP RPV steel SA508 Gr. 3

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongmin; Kim, Minchul; Choi, Kwonjae; Lee, Bongsang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this paper, the standard MC approach and BMC are applied to the forging material of the KSNP RPV steel SA508 Gr. 3. A series of fracture toughness tests were conducted in the DBTT transition region, and fracture toughness specimens were extracted from four regions, i.e., the surface, 1/8T, 1/4T and 1/2T. Deterministic material inhomogeneity was reviewed through a conventional MC approach and the random inhomogeneity was evaluated by BMC. In the present paper, four regions, surface, 1/8T, 1/4T and 1/2T, were considered for the fracture toughness specimens of KSNP (Korean Standard Nuclear Plant) SA508 Gr. 3 steel to provide deterministic material inhomogeneity and review the applicability of BMC. T0 determined by a conventional MC has a low value owing to the higher quenching rate at the surface as expected. However, more than about 15% of the KJC values lay above the 95% probability curves indexed with the standard MC T0 at the surface and 1/8T, which implies the existence of inhomogeneity in the material. To review the applicability of the BMC method, the deterministic inhomogeneity owing to the extraction location and quenching rate is treated as random inhomogeneity. Although the lower bound and upper bound curve of the BMC covered more KJC values than that of the conventional MC, there is no significant relationship between the BMC analysis lines and measured KJC values in the higher toughness distribution, and BMC and MC provide almost the same T0 values. Therefore, the standard MC evaluation method for this material is appropriate even though the standard MC has a narrow upper/lower bound curve range from the RPV evaluation point of view. The material is not homogeneous in reality. Such inhomogeneity comes in the effect of material inhomogeneity depending on the specimen location, heat treatment, and whole manufacturing process. The conventional master curve has a limitation to be applied to a large scatted data of fracture toughness such as the weld region

  8. Miniature Precracked Charpy Specimens for Measuring the Master Curve Reference Temperature of RPV Steels at Impact Loading Rates

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Scibetta, M.; Puzzolante, L.

    2008-10-15

    In the framework of the 2006 Convention, we investigated the applicability of fatigue precracked miniature Charpy specimens of KLST type (MPCC - B = 3 mm, W = 4 mm and L = 27 mm) for impact toughness measurements, using the well-characterized JRQ RPV steel. In the ductile to-brittle transition region, MPCC tests analyzed using the Master Curve approach and compared to data previously obtained from PCC specimens had shown a more ductile behavior and therefore un conservative results. In the investigation presented in this report, two additional RPV steels have been used to compare the performance of impact-tested MPCC and PCC specimens in the transition regime: the low-toughness JSPS steel and the high-toughness 20MnMoNi55 steel. The results obtained (excellent agreement for 20MnMoNi55 and considerable differences between T0 values for JSPS) are contradictory and do not presently allow qualifying the MPCC specimens as a reliable alternative to PCC samples for impact toughness measurements.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [Oak Ridge National Lab., TN (United States)] [and others

    1997-02-01

    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. Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany); Nikonov, S. [NRC ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2014-08-15

    The paper presents an uncertainty and sensitivity (U and S) study of the VVER-1000 reactor hydraulic properties. It is based on the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). The novelty of the work consists of taking into consideration all hydraulic uncertainty parameters used in the modeling of the reactor pressure vessel (RPV) internals. A detailed parallel channel ATHLET model of the RPV is developed. It consists of ca. 26 600 control volumes most of them connected with junctions for cross flows. The specific geometry of the gap between upper part of the baffle and upper part of fuel assembly and also a fuel assembly head is taken explicitly into account The influence of the input parameters on the sensitivity and uncertainty of the RPV outlet and inlet temperatures and mass flows as well assembly-wise mass flow and coolant temperature axial distributions is shown.

  11. Measurement of irradiation effects in a RPV steel by ball indentation technique and magnetic Barkhausen noise

    Science.gov (United States)

    Kim, In-Sup; Park, Duck-Gun; Byun, Thak-Sang; Hong, Jun-Hwa

    1999-12-01

    Effects of neutron dose on the mechanical and magnetic properties of a SA508-3 nuclear pressure vessel steel were investigated by using ball indentation test technique and magnetic Barkhausen noise (BN) measurements. The samples were irradiated in a research reactor up to 1018n/cm2 (E>1 MeV) at 70 °C. The yield strength and flow curve were evaluated from the indentation load-depth curves. The change of mechanical properties showed characteristic trend with respect to neutron dose, namely near plateau, rapid increase and slow increase. On the other hand, the BN varied in a reverse manner, a slow decrease up to a neutron dose of 1016n/cm2, followed by a rapid decrease up to a dose of 1018n/cm2.

  12. A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, Ernest D. [Modeling and Computing Services, LLC; Odette, George Robert [UCSB; Nanstad, Randy K [ORNL; Yamamoto, Takuya [ORNL

    2007-11-01

    The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance

  13. Development of High Strength and High Toughness Steels for Reactor Vessel and Surgeline Pipe

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. S.; Kim, M. C.; Yoon, J. H.; Kim, K. B.; Choi, K. J.; Cho, H. D.

    2010-07-15

    In addition to evaluating the effects of alloying elements, heat treatment conditions, weldability and neutron irradiation behavior were evaluated with 15 types of SA508 Gr.4N model alloys for reactor pressure vessel. The maximum yield strength of 630MPa were obtained by controlling chemical compositions and heat treatment conditions. Model alloys also showed excellent impact toughness and fracture toughness. The microstructure and mechanical properties of weld heat affected zone were evaluated by using simulated specimens and the effects of post weld heat treatment conditions were also investigated. Neutron irradiation behavior at high fluence level were characterized and then compared with commercial reactor pressure vessel steel. The value of transition temperature shift(TTS) was 22 .deg. C at 6.4x10{sup 19} n/cm{sup 2} which is similar to commercial RPV steel. However, its toughness after irradiation is much better than that of unirradiated commercial RPV steel due to the superior initial toughness. Leak-before-break(LBB) properties of type 316 stainless steel model alloys and their welds for surge line were evaluated as well as microstructure and mechanical properties. Tensile tests and J-R fracture resistance tests were carried out at RT and 316 .deg. C. The model alloys showed good tensile strength over standard value, except type 316L which has lower C/N. In the LBB safety analysis result, all of type 316 model alloys have higher allowable load than that of OPR1000 surge line

  14. Current understanding on the neutron irradiation embrittlement of BWR reactor pressure vessel steels in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Asano, K.; Nishiyama, T. [TEPCO (Japan); Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A. [CRIEPI (Japan); Ohta, T. [Japan Atomic Power Co. (Japan); Ishimaru, Y. [Chugoku EPCO (Japan); Yoneda, H. [Hokuriku EPCO (Japan); Lida, J. [Tohoku EPCO (Japan); Yuya, H. [Chubu EPCO (Japan)

    2011-07-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels has been of concern primarily for the pressurized water reactors (PWRs). After long operation experiences, we are now becoming aware of the situation that the neutron irradiation embrittlement is also of concern for some of the boiling water reactors (BWRs) particularly with Cu-containing RPV steels. The surveillance data of Cu-containing BWR RPV steels show relatively larger shift in ductile-to-brittle transition temperature of fracture toughness than predicted by the embrittlement correlation method developed in late eighties and early nineties. Accurate evaluation of the amount of embrittlement is now very important for long-term operation of BWRs. In this paper, we will describe the neutron irradiation embrittlement of BWR RPVs in Japan. Some of the materials that show relatively large transition temperature shifts are investigated to understand the causes of embrittlement using state-of-the-art microstructural characterization techniques. Furthermore, some archive materials of such RPVs are irradiated in a material testing reactor with high neutron flux to understand the effect of flux on transition temperature shifts and corresponding microstructural changes. Microstructural evolution under irradiation, solute clustering in particular could explain the differences in transition temperature shift of the analyzed specimens. Larger BWR RPVs, which have larger water gaps, receive less neutron irradiation and harmful impurities in steels such as copper are well controlled since 1980 so irradiation embrittlement in BWR vessels can now be considered a concern only in old and small plants. All the new information obtained through these activities was considered in the development of new embrittlement correlation that is now adopted in JEAC 4201- 2007 of Japan Electric Association

  15. Ductile-Brittle Transition Behavior in Tempered Martensitic SA508 Gr. 4N Ni-Mo-Cr Low Alloy Steels for Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Reactor pressure vessels (RPVs) operate under severe conditions of elevated temperature, high pressure, and irradiation. Therefore, a combination of sufficient strength, toughness, good weldability, and high irradiation resistance are required for RPV materials. SA508 Gr.4N low alloy steel, which has higher Ni and Cr contents than those of commercial RPV steel, Gr.3 steel, is considered as a candidate material due to its excellent mechanical properties from tempered martensitic microstructure. The ferritic steels such as Gr.3 and Gr.4N low alloy steels reveal a ductile-brittle transition and large scatters in the fracture toughness within a small temperature range. Recently, there are some observations of the steeper transition behavior in the tempered martensitic steels, such as Eurofer97 than the transition behavior of commercial RPV steels. It was also reported that the fracture toughness increased discontinuously when the phase fraction of the tempered martensite was over a critical fraction in the heat affected zones of SA508 Gr.3. Therefore, it may be necessary to evaluate the changes of transition behavior with a microstructure for the tempered martensitic SA508 Gr.4N low alloy steel. In this study, the fracture toughness for SA508 Gr.4N low alloy steels was evaluated from a view point of the temperature dependency with phase fraction of tempered martensite controlled by cooling rate. Additionally, a possible modification of the fracture toughness master curve was proposed and discussed

  16. 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)

  17. 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.

  18. 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.

  19. Microstructural dependence of Barkhausen noise and magnetic relaxation in the weld HAZ of an RPV steel

    Science.gov (United States)

    Park, Duck-Gun; Kim, Cheol Gi; Hong, Jun-Hwa

    2000-06-01

    Magnetic Barkhausen noise and permeability spectra have been measured to characterize different microstructure regions such as coarse-grain region, fine-grain region, intercritical structure (composed of tempered martensite and bainite) within the heat-affected zone (HAZ) of SA508-3 steel weldments using simulated HAZ microstructure sample. The intercritical region and coarse-grained region can be distinguished from the BNE and relaxation frequency. The BNE was decreased in the martensite regions and increased in the bainite regions by the post-weld heat treatment (PWHT). The change of relaxation frequency also showed similar trends, but the rate of change was less than that of BNE. The behavior of BNE and permeability spectra on the corresponding microstructure can be explained in terms of carbide morphology and residual stress related with domain wall motion.

  20. Master curve analysis of the SA508 Gr. 4N Ni-Mo-Cr low alloy steels for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Low alloy steels used as Reactor Pressure Vessels (RPVs) materials directly relate to the safety margin and the life span of reactors. Currently, SA508 Gr.3 low alloy steel is generally used for RPV material. But, for larger capacity and long-term durability of RPV, materials that have better properties including strength and toughness are needed. Therefore, tempered martensitic SA508 Gr.4N low alloy steel is considered as a candidate material due to excellent mechanical properties. The fracture toughness loss caused by irradiation embrittlement during reactor operation is one of the important issues for ferritic RPV steels, because the decrease of fracture toughness is directly related to the integrity of RPVs. One reliable and efficient concept to evaluate the fracture toughness of ferritic steels is master curve method. In ASTM E1921, it is clearly mentioned the universal shape of the median toughness-temperature curve for ferritic steels including tempered martensitic steels. However, currently, concerns have arisen regarding the appropriateness of the universal shape in ASTM for the tempered martensitic steels such as Eurofer97. Therefore, it may be necessary to assess the master curve applicability for the tempered martensitic SA508 Gr.4N low alloy steel. In this study, the fracture toughness behavior with temperature of the tempered martensitic SA508 Gr.4N low alloy steels was evaluated using the ASTM E1921 master curve method. And the results were compared with those of the bainitic SA508 Gr.3 low alloy steel. Furthermore, the way to define the fracture toughness behavior of Gr.4N steels well is discussed.

  1. Exploratory Study of Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A., Kryukov, A.M., Nikolaev, Y.A., Korolev, Y.N. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)], Sokolov, M.A., Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

    1997-12-31

    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 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; so far, only charpy impact and tensile specimens have been included. Oak Ridge National Laboratory (ornl) conducted such a program (irradiation and annealing) 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 (exp 19) N/SQ CM ({gt}1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 X 10 (exp 19) N/SQ CM ({gt}1 MeV).

  2. Study of a neutron irradiated reactor pressure vessel steel by X-ray absorption spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)], E-mail: sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2008-11-15

    Reactor pressure vessel (RPV) reference steel samples submitted to neutron irradiations followed by thermal annealing were investigated by X-ray absorption fine structure (XAFS) spectroscopy. Several studies revealed that Cu and Ni impurities can form nanoclusters. In the unirradiated sample and in the only-irradiated sample no significant clustering is detected. In all irradiated and subsequently annealed samples increases of Cu and Ni atom densities are recorded around the absorber. Furthermore, the density of Cu and Ni atoms determined in the first and second shells around the absorber is found to be affected by the irradiation and annealing treatment. The comparison of the XAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the local irradiation damage yields vacancy fractions which were determined from the analysis of XAFS data with a precision of {approx}5%.

  3. 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)

  4. Thermal Embrittlement of Reactor Pressure Vessel Steel due to Aging

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Soo; Park, Duck Gun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Thermal SS sets are located above the nuclear core where a fast neutron flux is negligible and temperature is 320 .deg. C (as opposed to 290 .deg. C in locations of high-irradiated SS). These SS allow monitoring of continuous operation temperature exposure effect on mechanical characteristics of the steels. Although transgranular cleavage is the predominant mode of brittle fracture in RPV steels, solute (e.g. phosphorus) segregation to grain boundaries can result in another type of brittle fracture known as intergranular (grain boundary) fracture. Figures 1 a) and b) show examples of transgranular and intergranular (IG) fracture, respectively, as viewed in a scanning electron microscope. The investigators have interpreted the intergranular cracking occurs as a result of segregation of sulfur and/or phosphorus at grain boundary. The IG cracking is a kind of symptom of embrittlement. It is reported that the IG cracking occurs in inert (Ar) environment under slow strain rate test. 1. The lath grain size in SA508 RPV steel increases slightly due to thermal aging at 350, 420, and 420 .deg. C for 2,250H. 2. The decrease in toughness appeared 4-25% and the lattice contraction appeared to be +0.004% - -0.022% due to thermal aging at 350, 420, and 420 .deg. C for 2,250H. 3. The amount of decrease in Charpy impact energy due to thermal aging is correlated well with the magnitude of lattice contraction.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-16

    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{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted

  7. A Review of the Application of Rate Theory to Simulate Vacancy Cluster Formation and Interstitial Defect Formation in Reactor Pressure Vessel Steel

    Directory of Open Access Journals (Sweden)

    Fallon Laliberte

    2015-10-01

    Full Text Available The beltline region of the reactor pressure vessel (RPV is subject to an extreme radiation, temperature, and pressure environment over several decades of operation; therefore it is necessary to understand the mechanisms through which radiation damage occurs and how it affects the mechanical and chemical properties of the RPV steel. Chemical rate theory is a mean field rate theory simulation model which applies chemistry to the evaluation of irradiation-induced embrittlement. It presents one method of analysis that may be coupled to other distinct methods, in order to analyze defect formation, ultimately providing useful information on strength, ductility, toughness and dimensional stability changes for effects such as embrittlement, reduction in ductility and toughness, void swelling, hardening, irradiation creep, stress corrosion cracking, etc. over time as materials are subjected to reactor operational irradiation. This paper serves as a brief review of rate theory fundamentals and presents several examples of research that exemplify the application and importance of rate theory in examining the effects of radiation damage on RPV steel.

  8. Heavy-Section Steel Irradiation Program

    Energy Technology Data Exchange (ETDEWEB)

    Rosseel, T.M.

    2000-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established.

  9. The changes of the structural, magnetic, and mechanical properties in a reactor pressure vessel steel neutron-irradiated at 70 .deg. C

    CERN Document Server

    Park, D G; Jang, K S; Jung, M M; Kim, G M

    1999-01-01

    The irradiation embrittlement of reactor-pressure-vessel steel has been one of the main safety concerns in nuclear power plants. In the present study, an SA508-3 RPV steel was irradiated by neutrons with various fluences up to 10 sup 1 sup 8 n/cm sup 2 (E>=1MeV) at a temperature of approximately 70 .deg. C. The irradiation responses of the structural, the magnetic, and the mechanical properties of the steel were investigated by means of X-ray diffraction, Moessbauer spectroscopy, magnetic Barkhausen noise, and micro-Vickers hardness measurements. The transitions of all of these parameters occurred above a neutron does of 10 sup 1 sup 6 n/cm sup 2. The results of the X-ray and the Moessbauer experiments revealed that neutron irradiation led to the possibility of partial amorphization in the investigated RPV steel. The changes of the physical and the mechanical properties were discussed in terms of irradiation-induced cascade damage of crystalline materials.

  10. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Maltsev, D. A.; Fedotova, S. V.; Frolov, A. S.; Zhuchkov, G. M.

    2017-01-01

    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

  11. Thermodynamic calculation and observation of microstructural change in Ni-Mo-Cr high strength low alloy RPV steels with alloying elements

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    An effective way of increasing the strength and fracture toughness of reactor pressure vessel steels is to change the material specification from that of Mn-Mo-Ni low alloy steel (SA508 Gr.3) to Ni-Mo-Cr low alloy steel(SA508 Cr.4N). In this study, we evaluate the effects of alloying elements on the microstructural characteristics of Ni-Mo-Cr low alloy steel. The changes in the stable phase of the SA508 Gr.4N low alloy steel with alloying elements were evaluated by means of a thermodynamic calculation conducted with the software ThermoCalc. The changes were then compared with the observed microstructural results. The calculation of Ni-Mo-Cr low alloy steels confirms that the ferrite formation temperature decreases as the Ni content increases because of the austenite stabilization effect. Consequently, in the microscopic observation, the lath martensitic structure becomes finer as the Ni content increases. However, Ni does not affect the carbide phases such as M{sub 23}C{sub 6} and M{sub 7}C{sub 3}. When the Cr content decreases, the carbide phases become unstable and carbide coarsening can be observed. With an increase in the Mo content, the M{sub 2}C phase becomes stable instead of the M{sub 7}C{sub 3} phase. This behavior is also observed in TEM. From the calculation results and the observation results of the microstructure, the thermodynamic calculation can be used to predict the precipitation behavior.

  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. 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.

  14. Development of High Strength Low Alloy Steel for Nuclear Reactor Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. S.; Kim, M. C.; Yoon, J. H; Choi, K. J.; Kim, J. M.; Hong, J. H.

    2013-11-15

    SA508 Gr. 4N Ni-Cr-Mo low alloy steel has an improved strength and fracture toughness, compared to commercial low alloy steels such as SA508 Gr. 3 Mn-Mo-Ni low alloy steel. In this study, the microstructural observation and baseline test were carried out using SA508 Gr. 4N model alloy of 1 ton scale. Thermal embrittlement and neutron irradiation embrittlement behaviors of SA508 Gr. 4N model alloy were also evaluated. The yield strength of 540MPa, Charpy transition temperature, T{sub 41J} of -132 .deg. C, Reference temperature, T{sub 0} of -146 .deg. C, and RT{sub NDT} of -105 .deg. C were obtained from large scale SA508 Gr. 3 low alloy steel. Effect of alloy elements on thermal embrittlement was carefully evaluated and embrittlement mechanism was characterized using small scale model alloys with various alloy composition. Neutron irradiation behavior at high fluence level up to 1.5x10{sup 20} n/cm{sup 2} corresponding over 80 years operation of RPV were investigated using irradiated samples from research reactor 'HANARO'. The irradiation embrittlement behavior of SA508 Gr. 4N model alloy was similar to that of commercial RPV steel. However, after neutron irradiation up to 1.3x10{sup 20} n/cm{sup 2}, SA508 Gr. 4N model alloy shows lower transition temperature(T{sub 41J} = -63 .deg. C) than unirradiated commercial RPV steel because it has a superior initial toughness.

  15. Radiation effects on reactor pressure vessel supports

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    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.

  16. Evaluation of Thermodynamic Stable Phase and Microstructure of SA508 Gr.4N Model Alloys for Reactor Pressure Vessel Steel with Variation of Alloying Elements

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mim Chul; Lee, B. S

    2009-12-15

    In order to increase the strength and the fracture toughness of RPV(reactor pressure vessel) steels, an effective way is the change of material specification from Mn-Mo-Ni low alloy steel(SA508 Gr.3) into Ni-Mo-Cr low alloy steel(SA508 Gr.4N). In this study, we evaluate the effects of alloying elements on microstructural characteristics in Ni-Mo-Cr low alloy steel. The changes in stable phase of SA508 Gr.4N low alloy steel with alloying elements were evaluated using a thermodynamic calculation by ThermoCalc software, and then compared with its microstructural observation results. From the calculation of Ni-Mo-Cr low alloy steels, ferrite formation temperature were decreased with increasing Ni and Mn contents due to austenite stabilization effect. Consequently, in the microscopic observation, the microstructure became finer with increasing Ni and Mn contents. However, they does not affects the carbide phase such as M{sub 23}C{sub 6} and M{sub 7}C{sub 3}. When the content of Cr is decreased, carbide phases became unstable and carbide coarsening is observed. With increase of Mo content, M{sub 2}C phase become stable instead of M{sub 7}C{sub 3} and it also observed in the TEM.

  17. 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.

  18. Continuous Cooling Transformations in Nuclear Pressure Vessel Steels

    Science.gov (United States)

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

    2014-10-01

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

  19. Studies on formation and structures of ultrafine Cu precipitates in Fe-Cu model alloys for reactor pressure vessel steels using positron quantum dot confinement in the precipitates by their positron affinity. JAERI's nuclear research promotion program, H11-034 (Contract research)

    CERN Document Server

    Hasegawa, M; Suzuki, M; Tang, Z; Yubuta, K

    2003-01-01

    Positron annihilation experiments on Fe-Cu model dilute alloys of nuclear reactor pressure vessel (RPV) steels have been performed after neutron irradiation in JMTR. Nanovoids whose inner surfaces were covered by Cu atoms were clearly observed. The nanovoids transformed to ultrafine Cu precipitates by dissociating their vacancies after annealing at around 400degC. The nanovoids and the ultrafine Cu precipitates are strongly suggested to be responsible for irradiation-induced embrittlement of RPV steels. Effects of Ni, Mn and P addition on the nanovoid and Cu precipitate formations were also studied. The nanovoid formation was enhanced by Ni and P, but suppressed by Mn. The Cu precipitates after annealing around 400degC were almost free from these doping elements and hence were pure Cu in the chemical composition. Furthermore the Fermi surface of the 'embedded' Cu precipitates with a body centered cubic crystal structure was obtained from two dimensional angular correlation of annihilation radiation (2D-ACAR) ...

  20. Normalizing treatment influence on the forged steel SAE 8620 fracture properties

    Directory of Open Access Journals (Sweden)

    Paulo de Tarso Vida Gomes

    2005-03-01

    Full Text Available In a PWR nuclear power plant, the reactor pressure vessel (RPV contains the fuel assemblies and reactor vessels internals and keeps the coolant at high temperature and high pressure during normal operation. The RPV integrity must be assured all along its useful life to protect the general public against a significant radiation liberation damage. One of the critical issues relative to the VPR structural integrity refers to the pressurized thermal shock (PTS accident evaluation. To better understand the effects of this kind of event, a PTS experiment has been planned using an RPV prototype. The RPV material fracture behavior characterization in the ductile-brittle transition region represents one of the most important aspects of the structural assessment process of RPV's under PTS. This work presents the results of fracture toughness tests carried out to characterize the RPV prototype material behavior. The test data includes Charpy energy curves, T0 reference temperatures for definition of master curves, and fracture surfaces observed in electronic microscope. The results are given for the vessel steel in the "as received" and normalized conditions. This way, the influence of the normalizing treatment on the fracture properties of the steel could be evaluated.

  1. Aspects of brittle failure assessment for RPV

    Energy Technology Data Exchange (ETDEWEB)

    Zecha, H.; Hermann, T.; Hienstorfer, W. [TUeV SUeD Energietechnik GmbH Baden-Wuerttemberg, Filderstadt (Germany); Schuler, X. [Materialpruefungsanstalt, Univ. Stuttgart (Germany)

    2009-07-01

    This paper describes the process of pressurized thermal shock analysis (PTS) and brittle failure assessment for the reactor pressure vessel (RPV) of the nuclear power plants NECKAR I/II. The thermo-hydraulic part of the assessment provides the boundary conditions for the fracture mechanics analysis. In addition to the one dimensional thermo-hydraulic simulations CFD, analyses were carried out for selected transients. An extensive evaluation of material properties is necessary to provide the input data for a reliable fracture mechanics assessment. For the core weld and the flange weld it has shown that brittle crack initiation can be precluded for all considered load cases. For the cold and hot leg nozzle detailed linear-elastic and elasticplastic Finite Element Analyses (FEA) are performed to verify the integrity of the RPV. (orig.)

  2. Steel Containment Vessel Model Test: Results and Evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

    A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. A concentric steel contact structure (CS), installed over the SCV model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. The SCV model and contact structure were instrumented with strain gages and displacement transducers to record the deformation behavior of the SCV model during the high pressure test. This paper summarizes the conduct and the results of the high pressure test and discusses the posttest metallurgical evaluation results on specimens removed from the SCV model.

  3. Preliminary results of steel containment vessel model test

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, T.; Komine, K.; Arai, S. [Nuclear Power Engineering Corp., Tokyo (Japan)] [and others

    1997-04-01

    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.

  4. Preliminary results of steel containment vessel model test

    Energy Technology Data Exchange (ETDEWEB)

    Luk, V.K.; Hessheimer, M.F. [Sandia National Labs., Albuquerque, NM (United States); Matsumoto, T.; Komine, K.; Arai, S. [Nuclear Power Engineering Corp., Tokyo (Japan); Costello, J.F. [Nuclear Regulatory Commission, Washington, DC (United States)

    1998-04-01

    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.

  5. Fatigue crack propagation in steels for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Klesnil, M.; Lukas, P.; Kunz, L. (Ceskoslovenska Akademie Ved, Brno. Ustav Fyzikalni Metalurgie); Troshchenko, V.T.; Pokrovskij, V.V.; Yasnij, P.V.; Skorenko, Y.S. (AN Ukrainskoj SSR, Kiev. Inst. Problem Prochnosti)

    1983-01-01

    Fatigue crack propagation data were measured on 15Kh2NMFA steel of Czechoslovak and Soviet makes. The results obtained by two laboratories were compared with other available data regarding materials for pressure vessels of nuclear power plants. Crack propagation curves were measured at temperatures -60, 20 and 350 degC and the corresponding parameters of crack growth equation were found. Threshold values of stress intensity factor amplitude, Ksub(apz), and the influence of stress ratio R in the range of small crack rates were determined experimentally. Fractography revealed either transgranular or mixed transgranular and interaranular fracture modes depending on stress intensity amplitude Ksub(a) and the environment.

  6. Effect of Heavy Ion Irradiation Dosage on the Hardness of SA508-IV Reactor Pressure Vessel Steel

    Directory of Open Access Journals (Sweden)

    Xue Bai

    2017-01-01

    Full Text Available Specimens of the SA508-IV reactor pressure vessel (RPV steel, containing 3.26 wt. % Ni and just 0.041 wt. % Cu, were irradiated at 290 °C to different displacement per atom (dpa with 3.5 MeV Fe ions (Fe2+. Microstructure observation and nano-indentation hardness measurements were carried out. The Continuous Stiffness Measurement (CSM of nano-indentation was used to obtain the indentation depth profile of nano-hardness. The curves showed a maximum nano-hardness and a plateau damage near the surface of the irradiated samples, attributed to different hardening mechanisms. The Nix-Gao model was employed to analyze the nano-indentation test results. It was found that the curves of nano-hardness versus the reciprocal of indentation depth are bilinear. The nano-hardness value corresponding to the inflection point of the bilinear curve may be used as a parameter to describe the ion irradiation effect. The obvious entanglement of the dislocations was observed in the 30 dpa sample. The maximum nano-hardness values show a good linear relationship with the square root of the dpa.

  7. Irradiation embrittlement of reactor pressure vessel steel at very high neutron fluence

    Science.gov (United States)

    Kryukov, A.; Debarberis, L.; von Estorff, U.; Gillemot, F.; Oszvald, F.

    2012-03-01

    For the prediction of radiation embrittlement of RPV materials beyond the NPP design time the analysis of research data and extended surveillance data up to a fluence ˜23 × 1020 cm-2 (E > 0.5 MeV) has been carried out. The experimental data used for the analysis are extracted from the International Database of RPV materials. Key irradiation embrittlement mechanisms, direct matrix damage, precipitation and element segregation have been considered. The essential part of the analysis concerns the assessment of irradiation embrittlement of WWER-440 steel irradiated with very high neutron fluence. The analysis of several surveillance sets irradiated at a fluence up to 23 × 1020 cm-2 (E > 0.5 MeV) has been performed. The effect of the main influencing chemical elements phosphorus and copper has been verified up to a fluence of 4.6 × 1020 cm-2 (E > 0.5 MeV). The data are indicating good radiation stability, in terms of the Charpy transition temperature shift and yield strength increase for steels with relatively low concentrations of copper and phosphorus. The linear dependence between ΔTk and ΔRp0.2 can be an evidence of strengthening mechanisms of irradiation embrittlement and absence of non-hardening embrittlement even at very high neutron fluence.

  8. Nonlinear Ultrasonic Techniques to Monitor Radiation Damage in RPV and Internal Components

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Kim, Jin-Yeon [Georgia Inst. of Technology, Atlanta, GA (United States); Qu, Jisnmin [Northwestern Univ., Evanston, IL (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wall, Joe [Electric Power Research Inst. (EPRI), Knoxville, TN (United States)

    2015-11-02

    The objective of this research is to demonstrate that nonlinear ultrasonics (NLU) can be used to directly and quantitatively measure the remaining life in radiation damaged reactor pressure vessel (RPV) and internal components. Specific damage types to be monitored are irradiation embrittlement and irradiation assisted stress corrosion cracking (IASCC). Our vision is to develop a technique that allows operators to assess damage by making a limited number of NLU measurements in strategically selected critical reactor components during regularly scheduled outages. This measured data can then be used to determine the current condition of these key components, from which remaining useful life can be predicted. Methods to unambiguously characterize radiation related damage in reactor internals and RPVs remain elusive. NLU technology has demonstrated great potential to be used as a material sensor – a sensor that can continuously monitor a material’s damage state. The physical effect being monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave. The degree of nonlinearity is quantified with the acoustic nonlinearity parameter, β, which is an absolute, measurable material constant. Recent research has demonstrated that nonlinear ultrasound can be used to characterize material state and changes in microscale characteristics such as internal stress states, precipitate formation and dislocation densities. Radiation damage reduces the fracture toughness of RPV steels and internals, and can leave them susceptible to IASCC, which may in turn limit the lifetimes of some operating reactors. The ability to characterize radiation damage in the RPV and internals will enable nuclear operators to set operation time thresholds for vessels and prescribe and schedule replacement activities for core internals. Such a capability will allow a more clear definition of reactor safety margins. The research consists of three tasks: (1

  9. 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..., peaks, bilges, machinery spaces, and bunkers shall be examined to determine the condition of the framing...

  10. 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.

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

    Science.gov (United States)

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

    2014-06-01

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

  12. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects; Mecanismes de fragilisation sous irradiation aux neutrons d'alliages modeles ferritiques et d'un acier de cuve: amas de defauts

    Energy Technology Data Exchange (ETDEWEB)

    Meslin-Chiffon, E

    2007-11-15

    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)

  13. Effect of Cooling Rate on Microstructures and Mechanical Properties in SA508 Gr4N High Strength Low Alloy Steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minchul; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The microstructure of Ni-Cr-Mo low alloy steel is a mixture of tempered martensite and tempered lower bainite and that of Mn-Mo-Ni low alloy steel is predominantly tempered upper bainite. Higher strength and toughness steels are very attractive as an eligible RPV steel, so several researchers have studied to use the Ni-Cr-Mo low alloy steel for the NPP application. Because of the thickness of reactor vessel, there are large differences in austenitizing cooling rates between the surface and the center locations of thickness in RPV. Because the cooling rates after austenitization determine the microstructure, it would affect the mechanical properties in Ni-Cr-Mo low alloy steel, and it may lead to inhomogeneous characteristics when the commercial scale of RPV is fabricated. In order to apply the Ni-Cr-Mo low alloy steel to RPV, it is necessary to evaluate the changes of microstructure and mechanical properties with varying phase fractions in Ni-Cr-Mo low alloy steel. In this study, the effects of martensite and bainite fractions on mechanical properties in Ni-Cr-Mo low alloy steel were examined by controlling the cooling rate after austenitization. First of all, continuous cooling transformation(CCT) diagram was established from the dilatometric analyses. Then, the phase fractions at each cooling rate were quantitatively evaluated. Finally, the mechanical properties were correlated with the phase fraction, especially fraction of martensite in Ni-Cr-Mo low alloy steel.

  14. Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang [Korea Atomic Energy Research Institute, Nuclear Material Research Div., Daejeon (Korea, Republic of)

    2012-10-15

    The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated.

  15. Effect of Silicon Content on Carbide Precipitation and Low-Temperature Toughness of Pressure Vessel Steels

    Science.gov (United States)

    Ruan, L. H.; Wu, K. M.; Qiu, J. A.; Shirzadi, A. A.; Rodionova, I. G.

    2017-05-01

    Cr - Mn - Mo - Ni pressure vessel steels containing 0.54 and 1.55% Si are studied. Metallographic and fractographic analyses of the steels after tempering at 650 and 700°C are performed. The impact toughness at - 30°C and the hardness of the steels are determined. The mass fraction of the carbide phase in the steels is computed with the help of the J-MatPro 4.0 software.

  16. The influence of the crust layer on RPV structural failure under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Jianfeng, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Li, Xiangqing [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Bao, Shiyi [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, Lijia [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Gao, Zengliang [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2017-05-15

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  17. Effects of Thermal Aging on Type 316H Stainless Steel for Reactor Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Hong, Seok Min; Lee, Bong Sang; Koo, Gyeong Hoi [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Type 316H stainless steel is a prime candidate for reactor vessel material of sodium-cooled fast reactor (SFR) which has been developed as one of the Gen IV nuclear reactors in Korea. The reactor vessel steel will be exposed to higher temperature for an extended design life time. It is known that austenitic stainless steel such as Type 316H stainless steel shows excellent toughness and adequate strength at a moderate temperature. However, the previous researches reported the mechanical properties of Type 316H stainless weld would be deteriorated by the aging at the elevated temperature range.

  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. Joining dissimilar stainless steels for pressure vessel components

    Science.gov (United States)

    Sun, Zheng; Han, Huai-Yue

    1994-03-01

    A series of studies was carried out to examine the weldability and properties of dissimilar steel joints between martensitic and austenitic stainless steels - F6NM (OCr13Ni4Mo) and AISI 347, respectively. Such joints are important parts in, e.g. the primary circuit of a pressurized water reactor (PWR). This kind of joint requires both good mechanical properties, corrosion resistance and a stable magnetic permeability besides good weldability. 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. The results of various tests indicated that the quality of the tube/tube joints is satisfactory for meeting all the design requirements.

  20. 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.

  1. Application of high strength MnMoNi steel to pressure vessels for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, K.; Kurihara, I.; Sasaki, T.; Koyama, Y.; Tanaka, Y. [The Japan Steel Works, Ltd. (Japan)

    1999-07-01

    Recent increase in output of nuclear power plant has been attained by enlargement of major components such as pressure vessels. Such large components have almost reached limit of size from the points of manufacturing capacity and cost in both forgemasters and fabricaters. In order to solve this problem, it must be beneficial to apply design by use of material of higher strength which brings reduction of pressure vessel thickness and weight. The Japan Steel Works, Ltd. (JSW) has many manufacturing experiences of large integrated forgings made from high strength MnMoNi steel with tensile strength level of 620MPa for steam generator (SG) pressure vessel, and has made confirmation tests of its material properties. This paper describes the confirmation test results such as tensile and impact properties, nil-ductility transition temperature (NDT-T), static and dynamic fracture toughness weldability including under clad cracking (UCC) sensitivity and metallurgical factors which influence on such material properties. (orig.)

  2. Application of high strength MnMoNi steel to pressure vessels for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, K. E-mail: koumei_suzuki@jsw.co.jp; Kurihara, I.; Sasaki, T.; Koyoma, Y.; Tanaka, Y

    2001-06-01

    Recent increase in output of nuclear power plant has been attained by enlargement of major components such as pressure vessels. Such large components have almost reached a size limit from the points of manufacturing capacity and cost in both forgemasters and fabricaters. In order to solve this problem, it must be beneficial to apply design by use of material of higher strength, which brings reduction of pressure vessel thickness and weight. The Japan Steel Works Ltd. (JSW) has many manufacturing experiences of large integrated forgings made from high strength MnMoNi steel with tensile strength level of 620 MPa for steam generator (SG) pressure vessel, and has performed confirmation tests of its material properties. This paper describes the confirmation test results such as tensile and impact properties, nil-ductility transition temperature (NDT-T), static and dynamic fracture toughness, weldability including under-clad cracking (UCC) sensitivity, as well as metallurgical factors which influence on such material properties.

  3. Hydrogen Cracking and Stress Corrosion of Pressure Vessel Steel ASTM A543

    Science.gov (United States)

    AlShawaf, Ali Hamad

    The purpose of conducting this research is to develop fundamental understanding of the weldability of the modern Quenched and Tempered High Strength Low Alloy (Q&T HSLA) steel, regarding the cracking behavior and susceptibility to environmental cracking in the base metal and in the heat affected zone (HAZ) when welded. A number of leaking cracks developed in the girth welds of the pressure vessel after a short time of upgrading the material from plain carbon steel to Q&T HSLA steel. The new vessels were constructed to increase the production of the plant and also to save weight for the larger pressure vessel. The results of this research study will be used to identify safe welding procedure and design more weldable material. A standardized weldability test known as implant test was constructed and used to study the susceptibility of the Q&T HSLA steel to hydrogen cracking. The charged hydrogen content for each weld was recorded against the applied load during weldability testing. The lack of understanding in detail of the interaction between hydrogen and each HAZ subzone in implant testing led to the need of developing the test to obtain more data about the weldability. The HAZ subzones were produced using two techniques: standard furnace and GleebleRTM machine. These produced subzones were pre-charged with hydrogen to different levels of concentration. The hydrogen charging on the samples simulates prior exposure of the material to high humidity environment during welding process. Fractographical and microstructural characterization of the HAZ subzones were conducted using techniques such as SEM (Scanning Electron Microscopy). A modified implant test using the mechanical tensile machine was also used to observe the effects of the hydrogen on the cracking behavior of each HAZ subzone. All the experimental weldability works were simulated and validated using a commercial computational software, SYSWELD. The computational simulation of implant testing of Q&T HSLA

  4. Studies on formation and structures of ultrafine Cu precipitates in Fe-Cu model alloys for reactor pressure vessel steels using positron quantum dot confinement in the precipitates by their positron affinity. JAERI's nuclear research promotion program, H11-034 (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Masayuki; Nagai, Yasuyoshi; Tang, Zheng; Yubuta, Kunio [Tohoku Univ., Sendai (Japan). Inst. for Materials Research; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Positron annihilation experiments on Fe-Cu model dilute alloys of nuclear reactor pressure vessel (RPV) steels have been performed after neutron irradiation in JMTR. Nanovoids whose inner surfaces were covered by Cu atoms were clearly observed. The nanovoids transformed to ultrafine Cu precipitates by dissociating their vacancies after annealing at around 400degC. The nanovoids and the ultrafine Cu precipitates are strongly suggested to be responsible for irradiation-induced embrittlement of RPV steels. Effects of Ni, Mn and P addition on the nanovoid and Cu precipitate formations were also studied. The nanovoid formation was enhanced by Ni and P, but suppressed by Mn. The Cu precipitates after annealing around 400degC were almost free from these doping elements and hence were pure Cu in the chemical composition. Furthermore the Fermi surface of the 'embedded' Cu precipitates with a body centered cubic crystal structure was obtained from two dimensional angular correlation of annihilation radiation (2D-ACAR) in a Fe-Cu single crystal and was agreed well with that from a band structure calculation. Theoretical calculation of positron confinement in Fe-Cu model alloys showed that a positron quantum dot state induced by positron affinity is attained for the embedded precipitates larger than 1 nm. A new position sensitive detector with a function of one dimensional angular correlation of annihilation radiation (1D-ACAR) has been developed that enables high resolution experiments over wide ranges of momentum distribution. (author)

  5. Investigation on multilayer failure mechanism of RPV with a high temperature gradient from core meltdown scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jianfeng, Mao, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China); Xiangqing, Li [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Shiyi, Bao, E-mail: bsy@zjut.edu.cn [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China); Lijia, Luo [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Zengliang, Gao [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China)

    2016-12-15

    Highlights: • The multilayer failure mechanism is investigated for RPV under CHF. • Failure time and location of RPV are predicted under various SA scenarios. • The structural behaviors are analyzed in depth for creep and plasticity. • The effect of internal pressure and temperature gradient is considered. • The structural integrity of RPV is secured within the required 72 creep hours. - Abstract: The Fukushima accident shows that in-vessel retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still exists inside the reactor pressure vessel (RPV). In the traditional concept of IVR, the pressure is supposed to successfully be released, and the temperature distributed among the wall thickness is assumed to be uniform. However, this concept is seriously challenged by reality of Fukushima accident with regard to the existence of both internal pressure and high temperature gradient. Therefore, in order to make the IVR mitigation strategy succeed, the numerical investigation of the lower head behavior and its failure has been performed for several internal pressures under high temperature gradient. According to some requirements in severe accident (SA) management of RPV, it should be ensured that the IVR mitigation takes effect in preventing the failure of the structure within a period of 72 h. Subsequently, the failure time and location have to be predicted under the critical heat flux (CHF) loading condition for lower head, since the CHF is limit thermal boundary before the melt-through of RPV. In illustrating the so called ‘multilayer failure mechanism’, the structural behaviors of RPV are analyzed in terms of the stress, creep strain, deformation, damage on selected paths.

  6. Study on probability of failure for RPV nozzle region under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Il Soon; Oh, Young Jin; Sim, Sang Hoon [Seoul National University, Seoul (Korea)

    2002-04-01

    Most of previous studies for creep rupture of RPV lower head under severe accident condition, have been focused on global failure of RPV lower head. In contrast, the local failure of the RPV nozzle region has not been studied in detail. This study focused the nozzle failure analysis into creep rupture evaluation of RPV lower head under severe accident condition, and this will help improve the safety assessment of nuclear power plants under severe accident conditions. The existence and features of nozzle failure in LAVA-ICI tested vessel of Korea Atomic Energy Research Institute and LHF-4 tested vessel of SNL, are examined. To understand the basic mechanical properties of nozzle material and weld metal, the tensile tests in various temperature levels and the creep rupture tests in various temperature and stress levels, are accomplished. The stress and deformation of LAVA-ICI experiments are analysed using measured basic mechanical properties. The failure time of Advanced Power Reactor 1400 (APR1400) in nozzle region was calculated using modified TMI-2 VIP model. Nozzle region failure characteristics was studied for SNL-LHF-4 experimental case using Finite Element Method (FEM). Using characteristics of nozzle failure, a new failure prediction experimental method was proposed for RPV nozzle failure. 19 refs., 43 figs. (Author)

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  9. Development of heavy steel plate for Mayflower Resolution, special purpose vessel for erection of offshore wind towers

    Energy Technology Data Exchange (ETDEWEB)

    Schuetz, W.; Schroeter, F.

    2005-05-15

    Special problems necessitate special solutions. Installation vessels for the erection of offshore wind towers are subject to extremely demanding design and structural specifications. Such projects are made possible only by the use of high strength, fine grained structural steels possessing good toughness properties even at extremely low temperatures; in addition, such steels must also offer good workability. Such steel plate material exhibits mechanical properties greatly superior to those possessed by conventional shipbuilding plate. This article focuses on the material for such an installation vessel and the underlying steel development work performed at AG der Dillinger Huettenwerke. (author)

  10. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M. [Nuclear Research Institute Rez plc (Czech Republic); Steele, L.E. [Chief Scientific Investigator of the Programme, Springfield, VA (United States)

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Eason, E.D.; Wright, J.E.; Nelson, E.E. [Modeling and Computing Services, Boulder, CO (United States); Odette, G.R.; Mader, E.V. [Univ. of California, Santa Barbara, CA (United States)

    1996-03-01

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

  12. Study on structural failure of RPV with geometric discontinuity under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Mao, J.F., E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Zhu, J.W. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Department of Mechanical and Electrical engineering, Huzhou Vocational & Technical College Huzhou, Zhejiang 313000 (China); Bao, S.Y., E-mail: bsy@zjut.edu.cn [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, L.J. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Gao, Z.L. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2016-10-15

    Highlights: • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage are the major contributor to RPV failure. • A elastic core is found at the midpoint of the highly-eroded region. • Weakest location has some ‘accommodating’ quality to prevent ductile tearing. • The internal pressure is critical for the determination of structural failure. - Abstract: A severe accident management strategy known as ‘in-vessel retention (IVR)’ is widely adopted in most of advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed period of time. This traditional concept of IVR without consideration of internal pressure effect wasn’t challenged until the occurrence of Fukushima accident on 2011, which showed that the structural behavior had not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still existed inside the RPV. Accordingly, the paper tries to address the related issue on whether lower head (LH) integrity can be maintained, when the LH is subjected to the thermal-mechanical loads created during such a severe accident. Because of the presence of the high temperature melt (∼1300 °C) on the inside of RPV, some local material is melted down to create a unique RPV with geometric discontinuity, while the outside of RPV submerged in cavity water will remain in nucleate boiling (at ∼150 °C). Therefore, the failure mechanisms of RPV can span a wide range of structural behaviors, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Through meticulous investigation, it is found that the RPV failure is mainly caused by creep and plasticity, especially for the inside of highly-eroded region. The elastic core (or layer) is found to exist in the proximity of mid-section of the highly-eroded wall. However, the elastic core is squeezed into

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

    Energy Technology Data Exchange (ETDEWEB)

    Eason, E.D.; Wright, J.E.; Nelson, E.E. [Modeling and Computing Services, Boulder, CO (United States); Odette, G.R.; Mader, E.V. [California Univ., Santa Barbara, CA (United States)

    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.

  14. Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States); Almirall, Nathan [Univ. of California, Santa Barbara, CA (United States); Robertson, Janet [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Server, W. L. [ATI Consulting, Pinehurst, NC (United States); Yamamoto, T. [Univ. of California, Santa Barbara, CA (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States)

    2017-05-01

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

  15. EFFECTS OF TEMPERING AND PWHT ON MICROSTRUCTURES AND MECHANICAL PROPERTIES OF SA508 GR.4N STEEL

    OpenAIRE

    Lee, Ki-Hyoung; JHUNG, MYUNG JO; Kim, Min-Chul; Lee, Bong-Sang

    2014-01-01

    Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT) conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV) material. The blocks of model alloy were austenitized at the conventional temperature of 880 °C, then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is signif...

  16. Effects of postweld heat treatment on the dissimilar weldments of SA508 Gr. 4N Ni-Mo-Cr low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Eun; Lee, Chang Hee [Hanyang University, Seoul (Korea, Republic of); Kim, Min Chul; Lee, Ho Jin; Kim, Keong Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jung, In Chul; Byeon, Jin Gwi [Doosan HEAVY Industries and Construction, Seoul (Korea, Republic of)

    2009-10-15

    Low alloy steels are generally employed in reactor pressure vessels (RPVs) of nuclear power plants. SA508 Gr.4N Ni-Mo-Cr low alloy steel has been studied as candidate materials for RPVs because it has excellent mechanical properties such as strength and toughness due to higher Ni and Cr contents compared with SA508 Gr.3 commercial RPV steel. In order to improve corrosion resistance of inner-wall of RPV, it is necessary to be coated by clad-welding with austenitic stainless steels such as 308L and 309L. PWHT of such dissimilar weldments results in the formation of a soft region near the weld interface on the low Cr side, and hard zone on the high Cr side of the weldment. The hardness change of weldments by formation of these characteristic zones affected overall safety margin and integrity property of RPV. In this study, effects of post-weld heat treatment (PWHT) on the microstructural and mechanical properties of dissimilar weldment between SA508 Gr.4N steel and austenitic stainless steel have been studied. The changes in microstructure, hardness and composition profiles across the weld interface were studied in detail. And, those results were compared with weldment in SA508 Gr.3 steel.

  17. 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.)

  18. 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.

  19. 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.

  20. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  1. Standard practice for examination of seamless, gas-filled, steel pressure vessels using angle beam ultrasonics

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice describes a contact angle-beam shear wave ultrasonic technique to detect and locate the circumferential position of longitudinally oriented discontinuities and to compare the amplitude of the indication from such discontinuities to that of a specified reference notch. This practice does not address examination of the vessel ends. The basic principles of contact angle-beam examination can be found in Practice E 587. Application to pipe and tubing, including the use of notches for standardization, is described in Practice E 213. 1.2 This practice is appropriate for the ultrasonic examination of cylindrical sections of gas-filled, seamless, steel pressure vessels such as those used for the storage and transportation of pressurized gasses. It is applicable to both isolated vessels and those in assemblies. 1.3 The practice is intended to be used following an Acoustic Emission (AE) examination of stacked seamless gaseous pressure vessels (with limited surface scanning area) described in Test Met...

  2. Effect on property of HIC-Resistance of vessel steel of PWHT

    Science.gov (United States)

    Zhao, Xinyu; Zou, Yang; Qin, Liye; Lv, Yanchun

    2017-09-01

    Post-Welding Heat Treatment (PWHT) is usually taken after welding and joining of vessel steel, which effects the mechanical and Hydrogen Induced Cracking (HIC) resistance property of vessel plates. Simulating PWHT experiment was taken to research on the effect on mechanical property and HIC-resistance of PWHT of vessel plates. Some conclusions can be summarized as following. Comparing with the normalizing samples, the tensile strength of the samples after PWHT with holding time of 2h, 6h and 20h decreases by 27MPa, 44MPa and 47MPa. Ductile Brittle Transition Temperature (DBTT) of the normalizing samples was almost close to those of the samples after PWHT. But the impact energy of samples at 0°C increased with rise of PWHT holding time. And the hardness of samples decreases with rise of PWHT holding time. As shown in morphology structure, the precipitation of carbonide increases with the rise of holding time of PWHT, which decreases the strength and hardness of samples and raises the impact energy. But PWHT has a little effect on HIC-Resistance, which means PWHT don't deteriorate property of HIC-Resistance of vessel plates severely.

  3. Cleavage Fracture Toughness of SA508 Gr.4N High Strength Low Alloy Steel with Different Phase Fraction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Kim, Min Chul; Choi, Kwon Jae; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Materials for reactor pressure vessel (RPV) are required to have good mechanical properties to endure the severe operating conditions inside the reactor. Various researches have focused on improving mechanical properties by the controlling the heat treatment process of commercial SA508 Gr.3 RPV steel. Some studies for identifying new material with high strength and toughness for larger capacity and longer lifetime of reactor are being performed. SA508 Gr.4N low alloy steel may be a promising RPV material due to its excellent mechanical properties from its tempered martensitic microstructure. Recently, some research showed that F/M steel composed of the tempered martensite has a steeper temperature dependency of the fracture toughness than the master curve expression. We have also focused on the steep transition properties of tempered martensitic SA508 Gr.4N steel in previous research. However, it has not yet confirmed that the transition behavior including temperature dependency with tempered martensite fraction. This investigation aims to evaluate the relationship between cleavage fracture toughness and tempered martensite fraction for SA508 Gr.4N low alloy steel. For this purpose, the model alloys were prepared by controlling the cooling rate from the austenitization temperature. The cleavage fracture toughness was characterized in transition temperature region by 3-point bending tests. Based on the test results and a stress distribution near crack tip calculated in FE analysis, the relationship between the carbide size distributions and the transition properties are analyzed

  4. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  5. 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.)

  6. Manufacturing Cost Analysis of Novel Steel/Concrete Composite Vessel for Stationary Storage of High-Pressure Hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Zhili [ORNL; Zhang, Wei [ORNL; Wang, Jy-An John [ORNL; Ren, Fei [ORNL

    2012-09-01

    A novel, low-cost, high-pressure, steel/concrete composite vessel (SCCV) technology for stationary storage of compressed gaseous hydrogen (CGH2) is currently under development at Oak Ridge National Laboratory (ORNL) sponsored by DOE s Fuel Cell Technologies (FCT) Program. The SCCV technology uses commodity materials including structural steels and concretes for achieving cost, durability and safety requirements. In particular, the hydrogen embrittlement of high-strength low-alloy steels, a major safety and durability issue for current industry-standard pressure vessel technology, is mitigated through the use of a unique layered steel shell structure. This report presents the cost analysis results of the novel SCCV technology. A high-fidelity cost analysis tool is developed, based on a detailed, bottom-up approach which takes into account the material and labor costs involved in each of the vessel manufacturing steps. A thorough cost study is performed to understand the SCCV cost as a function of the key vessel design parameters, including hydrogen pressure, vessel dimensions, and load-carrying ratio. The major conclusions include: The SCCV technology can meet the technical/cost targets set forth by DOE s FCT Program for FY2015 and FY2020 for all three pressure levels (i.e., 160, 430 and 860 bar) relevant to the hydrogen production and delivery infrastructure. Further vessel cost reduction can benefit from the development of advanced vessel fabrication technologies such as the highly automated friction stir welding (FSW). The ORNL-patented multi-layer, multi-pass FSW can not only reduce the amount of labor needed for assembling and welding the layered steel vessel, but also make it possible to use even higher strength steels for further cost reductions and improvement of vessel structural integrity. It is noted the cost analysis results demonstrate the significant cost advantage attainable by the SCCV technology for different pressure levels when compared to the

  7. A scaling and experimental approach for investigating in-vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R.E. [Fauske & Associates, Inc., Burr Ridge, IL (United States)

    1997-02-01

    The TMI-2 accident experienced the relocation of a large quantity of core material to the lower plenum. The TMI-2 vessel investigation project concluded that approximately 20 metric tonnes of once molten fuel material drained into the RPV lower head. As a result, the lower head wall experienced a thermal transient that has been characterized as reaching temperatures as high as 1100{degrees}C, then a cooling transient with a rate of 10 to 100{degrees}C/min. Two mechanisms have been proposed as possible explanations for this cooling behavior. One is the ingression of water through core material as a result of interconnected cracks in the frozen debris and/or water ingression around the crust which is formed on internal structures (core supports and in-core instrumentation) in the lower head. The second focuses on the lack of adhesion of oxidic core debris to the RPV wall when the debris contacts the wall. Furthermore, the potential for strain of the RPV lower head when the wall is overheated could provide for a significant cooling path for water to ingress between the RPV and the frozen core material next to the wall. To examine these proposed mechanisms, a set of scaled experiments have been developed to examine the potential for cooling. These are performed in a scaled system in which the high temperature molten material is iron termite and the RPV wall is carbon steel. A termite mass of 40 kg is used and the simulated reactor vessels have water in the lower head at pressures up to 2.2 MPa. Furthermore, two different thicknesses of the vessel wall are examined with the thicker vessel having virtually no potential for material creep during the experiment and the thinner wall having the potential for substantial creep. Moreover, the experiment includes the option of having molten iron as the first material to drain into the RPV lower head or molten aluminum oxide being the only material that drains into the test configuration.

  8. Effects of Phase Fraction on Temperature Dependency of Fracture Toughness in Transition Temperature Region in SA508 Gr. 4N Ni-Mo-Cr Low Alloy Steels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Wee, Dang Moon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The Reactor Pressure Vessel (RPV) is the main component in determining the lifetime of nuclear power plants because it is subject to the aging phenomenon of irradiation embrittlement and there is no practical method for replacing that component. For materials used for the RPV, sufficient strength and toughness are required to prevent failure against the severe operating conditions and the aging degradation of materials. SA508 Gr.4N Ni-Mo-Cr low alloy steel, in which Ni and Cr contents are higher than in conventional RPV steels, may be a promising RPV material with the improved strength and toughness from its tempered martensitic microstructure. Wallin observed that the temperature dependency of fracture toughness is not sensitive to the chemical composition, heat treatment, and irradiation for ferritic steels. This result led to the concept of a universal shape in the median toughness-temperature curve for all 'ferritic steels'. However, there are some doubts about the universal shape in the ASTM master curve for the tempered martensitic steels, such as Eurofer97. It was also reported that the fracture toughness increased discontinuously when the phase fraction of the tempered martensite was over a critical fraction in the heat affected zones of SA508 Gr.3. Therefore, it may be necessary to evaluate the changes of transition behavior with microstructures of steel. In this study, the effects phase fraction of tempered martensite controlled by a cooling rate on the transition behavior of SA508 Gr.4N low alloy steels was evaluated. Additionally, the relationship between the variations of yield strength with the temperature and fracture stress in a local approach was discussed

  9. 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.

  10. Mechanical spectroscopy of reactor-pressure-vessel steel embrittlement: a progress report

    Energy Technology Data Exchange (ETDEWEB)

    Van Ouytsel, K

    1998-08-01

    An enhanced surveillance strategy for testing the fracture toughness of reactor-pressure-vessel steel embrittlement is described. Microstructural investigation in support of damage modelling is an essential element in this enhanced strategy. Temperature-dependent experiments are very sensitive to differences in chemical composition and to effects of neutron irradiation as well as thermal ageing. Amplitude-dependent experiments can be related to tensile test results and correspond to a model for the yield stress. A full range of experiments were carried out on base and weld metal from the Doel-I-II power plants. The results have indicated that internal friction yields information which cannot always be detected by means of standard testing techniques. An inverted torsion pendulum for measuring internal friction has been constructed.

  11. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

    2009-01-01

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  12. Influence of fluence rate on radiation-induced mechanical property changes in reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R.; Hiser, A.L. (Materials Engineering Associates, Inc., Lanham, MD (USA))

    1990-03-01

    This report describes a set of experiments undertaken using a 2 MW test reactor, the UBR, to qualify the significance of fluence rate to the extent of embrittlement produced in reactor pressure vessel steels at their service temperature. The test materials included two reference plates (A 302-B, A 533-B steel) and two submerged arc weld deposits (Linde 80, Linde 0091 welding fluxes). Charpy-V (C{sub v}), tension and 0.5T-CT compact specimens were employed for notch ductility, strength and fracture toughness (J-R curve) determinations, respectively. Target fluence rates were 8 {times} 10{sup 10}, 6 {times} 10{sup 11} and 9 {times} 10{sup 12} n/cm{sup 2} {minus}s{sup {minus}1}. Specimen fluences ranged from 0.5 to 3.8 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The data describe a fluence-rate effect which may extend to power reactor surveillance as well as test reactor facilities now in use. The dependence of embrittlement sensitivity on fluence rate appears to differ for plate and weld deposit materials. Relatively good agreement in fluence-rate effects definition was observed among the three test methods. 52 figs., 4 tabs.

  13. 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.

  14. 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.

  15. 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.

  16. Characterization of Precipitation Behavior and Fracture Toughness along Thickness Direction in SA508 Gr.3 Mn-Mo-Ni low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jaemin; Kim, Min-Chul; Hong, Seokmin; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    SA 508 Gr.3 Mn-Mo-Ni low alloy steel forgings thicker than 200 mm are used for reactor pressure vessels in nuclear power plants. The cooling rate difference along the thickness direction during the quenching process causes variation in the microstructure and mechanical properties. The microstructural variation and its influence on the fracture toughness of RPV steels were investigated in this study. The cleavage fracture toughness in the transition region were evaluated with the ASTM E1921 master curve method for samples at different locations from the inner surface to the center thickness of the RPV steel. The microstructural features, such as the area fraction, and the size and distribution of precipitates were quantitatively evaluated at each sampling position. Microstructure observations showed that at near the surface position, bainite laths are finer, and furthermore, the carbides are smaller and homogeneously distributed. The fracture toughness at the surface was better than those at deeper positions. The reference temperature T{sub 0} showed a linear relationship with the area fraction of the carbides bigger than a certain critical size. It is concluded that the size of the precipitates caused by the cooling rate gradient may have a dominant role in controlling the cleavage fracture toughness variation along the thickness direction for a very thick RPV steel.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mizuno, R.; Matsuda, F. [NDE Center, Japan Power Engineering and Inspection Corp. (Japan); Brziak, P. [Welding Research Inst. - Industrial Inst. of Slovak Republic (Slovakia); Lomozik, M. [Inst. of Welding (Poland)

    2004-07-01

    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.)

  18. 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.

  19. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    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{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  20. Through-wall sampling of the Trawsfynydd RPV

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1996-06-01

    Four large, highly irradiated through-wall weld samples are to be removed from the Trawsfynydd Magnox reactor pressure vessel. The reactor was shut down in 1993 after 28 years of operation. The samples will be tested to investigate the integrity of steel pressure vessels. The choice of specialised tooling for the operation and its deployment are discussed. A Ultra High Power Pressure Water Jet cutting method has been selected to meet the demanding remote robotic requirements. (UK).

  1. Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C.E.; Raney, S.J. [comps.] [Oak Ridge National Lab., TN (United States)

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.

  2. Evaluation of the Temper embrittlement in SA508 Gr. 4N Low Alloy Steel with Ni, Cr Contents Variation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    It is well known that SA508 Gr.4N low alloy steel has an improved fracture toughness and strength, compared to commercial low alloy steels such as SA508 Gr.3 and SA533B which have less than 1% Ni. Higher strength and fracture toughness of low alloy steels could be achieved by Ni and Cr addition. So there are several researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature of a reactor pressure vessel is more than 300 .deg. C and it operates for over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, it requires phase stability in the high temperature range including temper embrittlement resistance. Although temper brittlement has not been reported in SA508 Gr.4N low alloy steel, the evaluation of the temper embrittlement phenomena on SA508 Gr.4N is required for an RPV application. In a previous study, we have concluded that additional Ni and Cr could change the microstructures of SA508 Gr.4N low alloy steel, and changed microstructure may affect the susceptibility of temper embrittlement in SA508 Gr.4N. In this study, we have performed a Charpy impact test of SA508 Gr.4N low alloy steel with changing alloying element contents such as Ni and Cr. The mechanical properties of these low alloy steels after a long-term heat treatment(450 .deg. C, 2000hr) are also evaluated. Then, the fracture modes of each impact specimens are examined and grain boundary segregation is analyzed by AES. The precipitation behaviors of the low alloy steels are observed by SEM.

  3. Study on the Segregation Behavior in SA508 Gr. 4N Low Alloy Steel with Mn Contents Variation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    It is generally known that SA508 Gr.4N low alloy steel has an improved fracture toughness and strength, compared to commercial low alloy steels such as SA508 Gr.3 and SA533B which have lower than 1% Ni. Higher strength and fracture toughness of low alloy steels could be achieved by adding the Ni and Cr. So there are several researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature of a reactor pressure vessel is more than 300 .deg. C and it operates for over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, it requires a phase stability in the high temperature range including temper embrittlement resistance. Although no temper embrittlement has been reported in SA508 Gr.4N low alloy steel, we need to evaluate the temper embrittlement phenomena on SA508 Gr.4N for an RPV application. In a previous study, we have concluded that additional Mn may accelerate the temper embrittlement effect in SA508 Gr.4N low alloy steel. So we need to examine the reason why Mn changes the susceptibility to temper embrittlement in SA508 Gr.4N. In this study, we have performed a Charpy impact test of SA508 Gr.4N low alloy steel at varing Mn contents. The mechanical properties of these low alloy steels after a long-term heat treatment(450 .deg. C, 2000hr) are evaluated. Then, the images of the fracture surfaces are observed and a grain boundary segregation is analyzed by AES and SIMS. We also analyze the grain boundary structures of the low alloy steels with EBSD.

  4. Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lucon, Enrico [National Inst. of Standards and Technology (NIST), Boulder, CO (United States)

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4 1019 n/cm2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 1013 n/cm2/s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 1013 n/cm2/s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 1019n/cm2. The irradiation-induced shifts of the Master Curve reference temperatures, ΔT0, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T0, 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT0, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  5. J-Integral characterization of the nozzle steels from intermediate test vessels IV-5 and IV-9

    Energy Technology Data Exchange (ETDEWEB)

    Auten, T.A.; Macdonald, B.D.; Scavone, D.W.; Bozik, D.

    1994-10-01

    Reported here are the results of elastic-plastic fracture toughness tests performed on low alloy steels from the nozzles of the intermediate test vessels IV-5 and IV-9 from the Heavy Steel Section Technology Program at Oak Ridge National Laboratory. These vessels had been given prototypic nozzle corner flaw tests prior to the development of the ASTM E-813 standard test procedure for J-integral testing. The objective of this work is to provide J-integral material test support for future elastic-plastic fracture mechanics analysis of the nozzles. J-integral tests at 88{degrees}C (190{degrees}F) of the IV-5 nozzle material produced stable ductile tearing. The tearing resistance data are expected to support analysis of the observed similar stable tearing response of the nozzle corner flaw. J-integral tests at 24{degrees}C (75{degrees}F) of the IV-9 nozzle produced elastic-plastic fracture instability preceded by stable tearing. A similar response was observed in the IV-9 nozzle corner flaw test. It will be a major and important challenge to develop a fracture mechanics rationale that reconciles these small specimen and nozzle corner flaw test results. These test results are being made available to allow their use by a wide variety of organizations in developing such a rationale, which would be a significant contribution to quantifying the flaw tolerance of reactor pressure vessels.

  6. Effects of Surface Roughness, Oxidation, and Temperature on the Emissivity of Reactor Pressure Vessel Alloys

    Energy Technology Data Exchange (ETDEWEB)

    King, J. L. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Jo, H. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Tirawat, R. [National Renewable Energy Laboratory, Concentrating Solar Power Group, Golden, Colorado; Blomstrand, K. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Sridharan, K. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin

    2017-08-31

    Thermal radiation will be an important mode of heat transfer in future high-temperature reactors and in off-normal high-temperature scenarios in present reactors. In this work, spectral directional emissivities of two reactor pressure vessel (RPV) candidate materials were measured at room temperature after exposure to high-temperature air. In the case of SA508 steel, significant increases in emissivity were observed due to oxidation. In the case of Grade 91 steel, only very small increases were observed under the tested conditions. Effects of roughness were also investigated. To study the effects of roughening, unexposed samples of SA508 and Grade 91 steel were roughened via one of either grinding or shot-peening before being measured. Significant increases were observed only in samples having roughness exceeding the roughness expected of RPV surfaces. While the emissivity increases for SA508 from oxidation were indeed significant, the measured emissivity coefficients were below that of values commonly used in heat transfer models. Based on the observed experimental data, recommendations for emissivity inputs for heat transfer simulations are provided.

  7. 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.

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

    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

  9. Mechanical properties of type 316L stainless steel welded joint for ITER vacuum vessel (1). Experiment of unirradiated welded joint

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Shigeru; Fukaya, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishiyama, Shintaro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takahashi, Hiroyuki; Koizumi, Kouichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-01-01

    In design activity of ITER, the vacuum vessel (VV) is ranked as one of the most important components in core reactor from the view point of first barrier to tritium release from the reactor. The VV of ITER is designed as double walled structure so that some parts of them are not qualified in the conventional design standards. So it is necessary to prepare the new design standards to be applied them. JAERI has executed the preparation activity of the new design standards and the technical data to support them. In this study, the results of metallographic observation and mechanical properties of unirradiated type 316L stainless steel welded joint were reported. (author)

  10. A study on the radiation damage and recovery of neutron irradiated vessel steel using magnetic Barkhausen noise

    Science.gov (United States)

    Park, Duck-Gun; Jeong, Hee-Tae; Hong, Jun-Hwa

    1999-04-01

    The radiation damage and thermal recovery characteristic of neutron irradiated SA508-3 reactor pressure vessel steel specimens have been investigated. Two recovery stages were identified from the results of hardness measurements during isochronal annealing and the mechanism responsible for the two stages was explained by using the results of Barkhausen noise measurement on the basis of the interaction between radiation induced defects and the magnetic domain wall. The coercivity was not changed by neutron irradiation, whereas the maximum magnetic induction increased. Barkhausen noise parameters associated with the domain wall motion were decreased by neutron irradiation and recovered with subsequent heat treatments.

  11. The effect of microstructural changes on magnetic barkhausen noise in Mn-Mo-Ni pressure vessel steel

    CERN Document Server

    Jeong, H T; Hong, J H; Ahn, Y S; Kim, G M

    1999-01-01

    The effect of microstructural changes on magnetic Barkhausen noise (BN) has been investigated in Mn-Mo-Ni pressure-vessel steel with various microstructures. The BN energy was strongly influenced by the microstructural features, such as the dislocation density, the residual stress, and the carbide morphology. The measured differences in BN signals are discussed on the basis of the domain wall dynamics associated with the microstructural states. The microstructures were observed by using atomic force microscopy(AFM), and the AFM results compared with the scanning electron microscopy observations.

  12. Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.B.; Bolton, C.J. [Magnox Electric plc, Berkeley Centre, Glos (United Kingdom)

    1997-02-01

    Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impurity precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.

  13. Estimation of radiation hardening in ferritic steels using the cluster dynamics models

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Jun Hyun; Kim, Whung Whoe; Hong, Jun Hwa [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    Evolution of microstructure under irradiation brings about the mechanical property changes of materials, of which the major concern is radiation hardening in this work. Radiation hardening is generally expressed in terms of an increase in yield strength as a function of radiation dose and temperature. Cluster dynamics model for radiation hardening has been developed to describe the evolution of point defects clusters (PDCs) and copperrich precipitates (CRPs). While the mathematical models developed by Stoller focus on the evolution of PDCs in ferritic steels under neutron irradiation, we slightly modify the model by including the CRP growth and estimate the magnitude of hardening induced by PDC and CRP. The model is then used to calculate the changes in yield strength of RPV steels. The calculation results are compared to measured yield strength values, obtained from surveillance testing of PWR vessel steels in France.

  14. 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.

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

    Science.gov (United States)

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

    1989-10-01

    The fatigue crack growth rate (FCGR) properties of SA508 C1 2a and SA533 Gr A C1 2 pressure vessel steels and the corresponding automatic submerged are weldments were developed in a high-temperature pressurized water (HPW) environment at 288 °C (550°F) and 7.2 MPa (1044 psi) at load ratios of 0.02 and 0.50. The HPW enviromment FCGR properties of these pressure vessel steels and submerged arc weldments were generally conservative, compared with the approrpriate American Society of Mechanical Engineers (ASME) Section XI water environmental reference curve. The growth rate of fatigue cracks in the base materials, however, was considerably faster in the HPW environment than in a corresponding 288°C (550°F) base line air environment. The growth rate of fatigue cracks in the two submerged are weldments was also accelerated in the HPW environment but to a significantly lesser degree than that demonstrated by the corresponding 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, as compared with the weldments, was attributed to a different sulfide composition and morphology.

  16. J-R fracture characteristics of ferritic steels for RPVs and RCS piping of nuclear power plants

    Science.gov (United States)

    Yoon, Ji-Hyun; Lee, Bong-Sang; Hong, Jun-Hwa

    2001-10-01

    J-R fracture resistance tests have been performed on 3 heats of SA508-Gr.3 nuclear reactor pressure vessel (RPV) steel as well as 2 heats of SA516-Gr.70 and a heat of SA508-Gr.1a steels for nuclear reactor coolant system (RCS) piping. For the latter two steels, dynamic in addition to static J-R fracture resistances were investigated. From the test results of the SA508-Gr.3 steels, the J-R fracture resistance was superior in the following order: Si-killing steel, modified VCD steel and VCD steel. Microstructural analyses were carried out to correlate J-R fracture resistances with microstructural characteristics. According to the test results for SA508-Gr.1a and SA516-Gr.70 steels, all of the tested steels showed steep drops in fracture resistance at certain temperature and loading rate combinations. One heat of SA516-Gr.70 steel was very sensitive to dynamic strain aging and its fracture resistance was significantly low. It was concluded that microstructural and chemical factors affect the J-R fracture and DSA characteristics of SA516-Gr.70 steels.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  18. Reactor pressure vessel. Status report

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  19. Mechanical properties of a modified 2 1/4 Cr-1 Mo steel for pressure vessel applications. [V-Ti-B-modified

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Swindeman, R.W.

    1983-12-01

    Tensile and creep properties were determined on a V-Ti-B-modified 2 1/4 Cr-1 Mo steel considered to be a candidate alloy for pressure vessel aplications for coal liquefaction. The modified 2 1/4 Cr-1 Mo steel had about 0.2% V added for improved elevated-temperature strength and 0.02% Ti for grain refinement. Boron was added to improve the hardenability, thus allowing thicker sections to be quenched and normalized to completely bainitic microstructures. Lower carbon and silicon concentrations were used (approx. 0.1% C and 0.02% Si) than in standard 2 1/4 Cr-1 Mo steel. The mechanical properties determined on the modified steel after a heat treatment typical for SA-387, grade 22, class 2, indicated high toughness and excellent elecated-temperature tensile and creep strength. The modified steel had substantially better stress-rupture properties than did a standard 2 1/4 Cr-1 Mo steel (both with bainitic microstructures) with equivalent tensile properties - especially at the lowest stresses and highest temperatures. The modified steel had toughness properties superior to those of the standard 2 1/4 Cr-1 Mo steel. Comparative transmission electron microscopy studies of the standard and modified 2 1/4 Cr-1 Mo steels indicated that the differences involve the carbide precipitates and the dislocation substructures present in the steels.

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

    OpenAIRE

    Yang, Zhi-Qiang; Liu, Zheng-Dong; HE Xi-kou; Liu, Ning

    2017-01-01

    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...

  1. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-15

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

  2. 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.

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

    Science.gov (United States)

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

    2011-11-01

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

  4. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, Vladimír, E-mail: vladimir.slugen@stuba.sk; Pecko, Stanislav; Sojak, Stanislav

    2016-01-15

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1–2 vacancies with relatively small contribution (with intensity on the level of 20–40 %) were observed in “as-received” steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2–3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  5. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Science.gov (United States)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  6. 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)

  7. Structural integrity of nuclear reactor pressure vessels

    Science.gov (United States)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  8. 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.

  9. Plastic deformation and fracture behaviour of 21/4 Cr-1 Mo pressure-vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Holzmann, M.; Vlach, B.; Man, J.; Bilek, Z.

    1989-01-01

    During the heat treatment of steel plates and forgings of large thicknesses, microstructures with various volume fractions of ferrite appear. Plastic properties and fracture behaviour of these mixed microstructures are a function of ferrite content. The influence of ferrite content in the range from 0% to 54% in the bainitic-ferritic microstructure on mechanical properties and fracture behaviour of 21/4 Cr-1 Mo steel was examined. The yield stress was found to decrease linearly with the volume fraction of ferrite. The tensile strength was independent of ferrite content up to 25%, after which the tensile strength decreased. Using the Charpy test it has been found that the critical ferrite content-25%-exists in a mixed microstructure, at which the propagation and initiation transition temperatures attain the highest values. The fracture toughness tests gave the same results. Increasing the volume fraction of ferrite, the cleavage fracture toughness/temperature curves were shifted to higher temperatures. Simultaneously, the ductile-brittle fracture toughness transition temperature was raised reaching the highest value for the critical ferrite content. The fracture behaviour could be tentatively explained through the influence of ferrite volume fraction on both the cleavage fracture stress and the stress level at the crack tip.

  10. Estimation of lower-bound K{sub Jc} on pressure vessel steels from invalid data

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

    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{sub Ic} or K{sub 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{sub 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{sub 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.

  11. Weldability and toughness evaluation of pressure vessel quality steel using the shielded metal arc welding (SMAW) process

    Science.gov (United States)

    Datta, R.; Mukerjee, D.; Mishra, S.

    1998-12-01

    The present study was carried out to assess the weldability properties of ASTM A 537 Cl. 1 pressure-vessel quality steel using the shielded metal arc welding (SMAW) process. Implant and elastic restraint cracking (ERC) tests were conducted under different welding conditions to determine the cold cracking susceptibility of the steel. The static fatigue limit values determined for the implant test indicate adequate resistance to cold cracking even with unbaked electrodes. The ERC test, however, established the necessity to rebake the electrodes before use. Lamellar tearing tests carried out using full-thickness plates under three welding conditions showed no incidence of lamellar tearing upon visual examination, ultrasonic inspection, and four-section macroexamination. Lamellar tearing tests were repeated using machined plates, such that the central segregated band located at the midthickness of the plate corresponded to the heat-affected zone (HAZ) of the weld. Only in one (no rebake, heat input: 14.2 kj cm-1, weld restraint load: 42 kg mm-2) of the eight samples tested was lamellar tearing observed. This was probably accentuated due to the combined effects of the presence of localized pockets of a hard phase (bainite) and a high hydrogen level (unbaked electrodes) in the weld joint. Optimal welding conditions were formulated based on the above tests. The weld joint was subjected to extensive tests and found to exhibit excellent strength (tensile strength: 56.8 kg mm-2, or 557 MPa), and low temperature impact toughness (7.4 and 4.5 kg-m at-20 °C for weld metal, WM, and HAZ) properties. Crack tip opening displacement tests carried out for the WM and HAZ resulted in δm values 0.36 and 0.27 mm, respectively, which indicates adequate resistance to brittle fracture.

  12. Fracture toughness and the master curve for modified 9Cr-1Mo steel

    Science.gov (United States)

    Yoon, Ji-Hyun; Yoon, Eui-Pak

    2006-12-01

    Modified 9Cr-1Mo steel is a primary candidate material for the reactor pressure vessel of a Very High Temperature Gas-Cooled Reactor (VHTR) in the Korean Nuclear Hydrogen Development and Demonstration (NHDD) program. In this study, the T0 reference temperature, J-R fracture resistance and Charpy impact properties were evaluated for commercial Grade 91 steel as part of the preliminary testing for a selection of the RPV material for the VHTR. The fracture toughness of the modified 9Cr-1Mo steel was compared with that of SA508-Gr.3. The objective of this study was to obtain the pre-irradiation fracture toughness properties of the modified 9Cr-1Mo steel as reference data for an investigation of radiation effects. Charpy impact properties of the modified 9Cr-1Mo steel were similar to those of SA508-Gr.3. T0 reference temperatures were measured as -67.7 and -72.4°C from the tests with standard PCVN (pre-cracked Charpy V-notch) and half-sized PCVN specimens respectively, which were similar to the results for SA508-Gr.3. The KJc values of the modified 9Cr-1Mo steel with the test temperatures are successfully expressed by the Master Curve. The J-R fracture resistance of the modified 9Cr-1Mo steel at room temperature was nearly identical to that of SA508-Gr.3; in contrast, it was slightly higher at an elevated temperature.

  13. 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.

  14. Effects of alloying element contents on the toughness and transition behavior in the SA508 Gr. 4N Ni-Mo-Cr low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Park, Sang Gyu; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Low alloy steels used as materials for reactor pressure vessels (RPVs) determine the safety and the life span of reactors. Currently, SA508 Gr.3 low alloy steel is generally used for RPV materials. But, for larger capacity and long-term durability of the next generation RPVs, materials that have much better properties are needed, such as strength, toughness and irradiation resistance. SA508 Gr.4N low alloy steel shows good mechanical properties due to high Ni and Cr contents in comparison with the currently used reactor pressure vessel steels. Materials for RPVs suffer a decrease of toughness due to an embrittlement of the materials by neutron irradiation, especially in ferritic steels. This toughness loss causes an increase in the transition temperature, and then a brittle fracture could occur. Therefore, for an integrity assessment of low alloy steels as RPVs, an accurate evaluation of the transition behavior is needed, such as fracture and impact toughness. In this study, the toughness and transition behavior of SA 508 Gr.4N low alloy steels, which have different Ni, Cr and Mo, were evaluated in the transition region. And the applicability of the test results for Master-Curve method was assessed. Additionally, differences between influences of alloying elements contents on Charpy impact toughness and fracture toughness were discussed in terms of microstructural features.

  15. 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.

  16. Fatigue crack growth rates in a pressure vessel steel under various conditions of loading and the environment

    Science.gov (United States)

    Hicks, P. D.; Robinson, F. P. A.

    1986-10-01

    Corrosion fatigue (CF) tests have been carried out on SA508 Cl 3 pressure vessel steel, in simulated P.W.R. environments. The test variables investigated included air and P.W.R. water environments, frequency variation over the range 1 Hz to 10 Hz, transverse and longitudinal crack growth directions, temperatures of 20 °C and 50 °C, and R-ratios of 0.2 and 0.7. It was found that decreasing the test frequency increased fatigue crack growth rates (FCGR) in P.W.R. environments, P.W.R. environment testing gave enhanced crack growth (vs air tests), FCGRs were greater for cracks growing in the longitudinal direction, slight increases in temperature gave noticeable accelerations in FCGR, and several air tests gave FCGR greater than those predicted by the existing ASME codes. Fractographic evidence indicates that FCGRs were accelerated by a hydrogen embrittlement mechanism. The presence of elongated MnS inclusions aided both mechanical fatigue and hydrogen embrittlement processes, thus producing synergistically fast FCGRs. Both anodic dissolution and hydrogen embrittlement mechanisms have been proposed for the environmental enhancement of crack growth rates. Electrochemical potential measurements and potentiostatic tests have shown that sample isolation of the test specimens from the clevises in the apparatus is not essential during low temperature corrosion fatigue testing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 10{sup 19} n cm{sup −2} (E > 1 MeV) and a flux of 1.1 × 10{sup 13} n cm{sup −2} s{sup −1} at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

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

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Suzuki, M. [Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan)

    2014-09-15

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

  19. On the role of sulfur on the dissolution of pressure vessel steels at the tip of a propagating crack in PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Combrade, P.; Foucault, M. (UNIREC 42- Firminy (FR)); Marcus, P. (Ecole Nationale Superieure de Chimie 75 - Paris (FR)); Slama, G. (Societe Franco-Americaine de Constructions Atomiques (Framatome), 92 - Courbevoie (FR))

    1990-03-01

    Different aspects of the effect of sulfur on the dissolution and film repair on pressure vessel steel exposed to PWR environment at 300{sup 0}C were examined. A monolayer of sulfur adsorbed on a bare surface was shown to inhibit the nucleation of a magnetite film. The comparison of this result with dissolution measurements performed by using CERT under controlled potential lead to the assumption that mechanical rupture steps are involved in the environmental effect on the crack propagation rate. 27 refs.

  20. Effect of heterogeneities on the thermoelectric power of pressure vessel steel; Effet des heterogeneites sur le pouvoir thermoelectrique de l'acier de cuve

    Energy Technology Data Exchange (ETDEWEB)

    Simonet, L

    2006-12-15

    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)

  1. The environmentally-assisted cracking behaviour in the transition region of nickel-base alloy/low-alloy steel dissimilar weld joints under simulated BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P.; Leber, H.J. [Paul Scherrer Institute, Nuclear Energy and Safety Research Department, Lab for Nuclear Materials, 5232 Villigen PSI (Switzerland)

    2011-07-01

    The stress corrosion cracking (SCC) behaviour perpendicular to the fusion line in the transition region between the Alloy 182 nickel-base weld metal and the adjacent low-alloy reactor pressure vessel (RPV) steel of simulated dissimilar metal weld joints was investigated under boiling water reactor normal water chemistry conditions at different stress intensities and chloride concentrations. A special emphasis was placed to the question whether a fast growing inter-dendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of inter-dendritic stress corrosion crack growth was observed in high-purity or sulphate-containing oxygenated water under periodical partial unloading or constant loading conditions with stress intensity factors below 60 MPa-m{sup 1/2} for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the inter-dendritic stress corrosion crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high growth rate as a transgranular crack into the heat-affected zone and base material of the adjacent low-alloy steel. (authors)

  2. SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, William M.; Riley, Matthew E.; Spencer, Benjamin W.

    2016-07-01

    In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vessels – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in

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

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL; Terry, Totemeier [Idaho National Laboratory (INL)

    2006-10-01

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

  4. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

    1999-12-01

    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  5. Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho [KAERI, Daejeon (Korea, Republic of); Lee, Ki-Hyoung [KAIST, Daejeon (Korea, Republic of); Lee, Chang-Hee [Hanyang Univ., Seoul (Korea, Republic of)

    2011-08-15

    SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

  6. Main results of study on the interaction between the corium melt and steel in the VVER-1000 reactor vessel during a severe accident performed under the MASCA project

    Science.gov (United States)

    Asmolov, V. G.; Zagryazkin, V. N.; Tsurikov, D. F.; Vishnevsky, V. Yu.; D'Yakov, Ye. K.; Kotov, A. Yu.; Repnikov, V. M.

    2010-12-01

    The interactions that take place in the corium melt in the reactor vessel in the case of a severe accident at a nuclear power plant were investigated in accordance with the MASCA international program. Results of the interaction between the oxide melt and iron (steel), partition of the main components [U, Zr, Fe (stainless steel)] between the oxide and the metal phases of the melt, partition of low-volatile simulators of fission products between the phases of the stratified core melt pool, and impact of the oxidizing atmosphere on the melt stratification are presented. The results obtained were used for prediction of thermodynamic properties of the melts belonging to the U-Zr-Fe-O system.

  7. The efficacy, pharmacokinetics, safety and cardiovascular risks of switching nevirapine to rilpivirine in HIV-1 patients: the RPV switch study

    NARCIS (Netherlands)

    Rokx, C.; Blonk, M.; Verbon, A.; Burger, D.M.; Rijnders, B.J.

    2014-01-01

    INTRODUCTION: Nevirapine (NVP) induces cytochrome P450 3A4 by which rilpivirine (RPV) is metabolized. Switching NVP to RPV could result in decreased RPV exposure with subsequent virological failure and dyslipidemia because NVP is regarded as the least dyslipidemic, non-nucleoside, reverse

  8. The role of water chemistry for environmentally assisted cracking in low-alloy reactor pressure vessel and piping steels under boiling reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2005-07-01

    The environmentally assisted initiation and propagation of cracks in structural materials is one of the most important degradation and ageing mechanisms in light water reactors (LWR) and may seriously affect plant availability and economics. In the first part of this paper a short general introduction on environmentally assisted cracking (EAC) and its significance for LWR is given. Then the important role of water chemistry control in reducing the EAC risk in LWR is illustrated by current research results about the effect of chloride transients and hydrogen water chemistry on the EAC crack growth behaviour of low-alloy reactor pressure vessel and piping steels under boiling water reactor conditions. (author)

  9. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  10. Investigation of the Effects of Submerged Arc Welding Process Parameters on the Mechanical Properties of Pressure Vessel Steel ASTM A283 Grade A

    Directory of Open Access Journals (Sweden)

    Prachya Peasura

    2017-01-01

    Full Text Available The pressure vessel steel is used in boilers and pressure vessel structure applications. This research studied the effects of submerged arc welding (SAW process parameters on the mechanical properties of this steel. The weld sample originated from ASTM A283 grade A sheet of 6.00-millimeter thickness. The welding sample was treated using SAW with the variation of three process factors. For the first factor, welding currents of 260, 270, and 280 amperes were investigated. The second factor assessed the travel speed, which was tested at both 10 and 11 millimeters/second. The third factor examined the voltage parameter, which was varied between 28 and 33 volts. Each welding condition was conducted randomly, and each condition was tested a total of three times, using full factorial design. The resulting materials were examined using tensile strength and hardness tests and were observed with optical microscopy (OM and scanning electron microscopy (SEM. The results showed that the welding current, voltage, and travel speed significantly affected the tensile strength and hardness (P value < 0.05. The optimum SAW parameters were 270 amperes, 33 volts, and 10 millimeters/second travel speed. High density and fine pearlite were discovered and resulted in increased material tensile strength and hardness.

  11. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  12. Overview of the RPV-2 and INTERN-1 packages: From primary damage to microplasticity

    Science.gov (United States)

    Adjanor, G.; Bugat, S.; Domain, C.; Barbu, A.

    2010-11-01

    In the framework of the European project PERFECT, four multiscale simulation packages dedicated to the prediction of evolution of material properties were developed. Among them, the RPV-2 and INTERN-1 are two simulation sequences of similar structure dealing with radiation damage in the reactor pressure vessel and the reactor internal structures, respectively. Both start at the atomic scale, where the neutron spectrum of the specified reactor is used to determine the energy distribution of the primary knocked-on atoms (PKA). A database of molecular dynamics results is then used to integrate the instantaneous production of defect clusters resulting from the displacement cascades initiated by each PKA. Depending on the type of calculation chosen to model long-term diffusion and reactions of defect clusters, precipitates and mixed-clusters, this primary damage enters either in rate equations or in Object Kinetic Monte Carlo simulations. The later correspond to a more accurate (but also more computationally demanding) physical model for diffusion as positions of objects on a lattice are explicitly treated. Finally, the increase of critical resolved shear stress is estimated from these cluster distributions either using an analytical model, taking into account the self and mutual dipole interactions of dislocations pinned on randomly dispersed unshearable obstacles, or by simulating the glide of a single dislocation line in its main slip system. Dislocation dynamics simulations were already used to validate some of the assumptions of the latter models, and will be fully integrated in the next versions of the packages.

  13. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  14. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, Dresden (Germany)

    2009-07-01

    Russian type WWER reactors are operated in Russia and many other European countries like Finland, Czech Republic, Slovak Republic, Hungary, Bulgaria and Ukraine. Surveillance specimen programmes for the inspection of the aging of the reactor pressure vessel (RPV) materials were implemented for the second generation of WWER-440/V-213 reactors. The test results and the RPV integrity assessment has been evaluated according to national codes based on the Russian code PNAE G-7-002-86 ''Strength Calculation Norms for Nuclear Power Plant Equipment and Piping'' [1]. This is an indirect and correlative approach of determining the fracture toughness of the RPV steels in the initial and irradiated condition. The Master Curve (MC) approach as adopted in the test procedure ASTM E1921 [2] for assessing the fracture toughness of sampled irradiated materials has been gaining acceptance throughout the world [3]. The MC approach is more naturally suited to probabilistic analyses because it defines both a mean transition toughness value and a distribution around that value. It contains the assumptions of macroscopically homogenous material with uniform distribution of crack initiating defects along the crack front. In contrast to present indirect and correlative approach the specimen orientation and especially the crack extension direction in multilayer weld metal becomes more important for the direct measurement of the fracture toughness with Charpy size SE(B) specimens. The orientation of the Charpy- and SE(B) specimens is different for RPVs manufactured in Russia and by the SKODA company in the former Czechoslovakia [4,5]. Particularly with regard to weld metal it can be expected that the parameters of fracture toughness measured with Charpy-V or SE(B) specimens are strongly influenced by the specimen orientation. It raises the question whether the MC approach is also applicable when the structure varies along the crack front which is happen in TL oriented SE

  15. Master curve characterization of the fracture toughness behavior in SA508 Gr.4N low alloy steels

    Science.gov (United States)

    Lee, Ki-Hyoung; Kim, Min-Chul; Lee, Bong-Sang; Wee, Dang-Moon

    2010-08-01

    The fracture toughness properties of the tempered martensitic SA508 Gr.4N Ni-Mo-Cr low alloy steel for reactor pressure vessels were investigated by using the master curve concept. These results were compared to those of the bainitic SA508 Gr.3 Mn-Mo-Ni low alloy steel, which is a commercial RPV material. The fracture toughness tests were conducted by 3-point bending with pre-cracked charpy (PCVN) specimens according to the ASTM E1921-09c standard method. The temperature dependency of the fracture toughness was steeper than those predicted by the standard master curve, while the bainitic SA508 Gr.3 steel fitted well with the standard prediction. In order to properly evaluate the fracture toughness of the Gr.4N steels, the exponential coefficient of the master curve equation was changed and the modified curve was applied to the fracture toughness test results of model alloys that have various chemical compositions. It was found that the modified curve provided a better description for the overall fracture toughness behavior and adequate T0 determination for the tempered martensitic SA508 Gr.4N steels.

  16. Master curve characterization of the fracture toughness behavior in SA508 Gr.4N low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki-Hyoung, E-mail: shirimp@kaist.ac.k [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of); Kim, Min-Chul; Lee, Bong-Sang [Nuclear Materials Research Division, KAERI, Daejeon 305-353 (Korea, Republic of); Wee, Dang-Moon [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of)

    2010-08-15

    The fracture toughness properties of the tempered martensitic SA508 Gr.4N Ni-Mo-Cr low alloy steel for reactor pressure vessels were investigated by using the master curve concept. These results were compared to those of the bainitic SA508 Gr.3 Mn-Mo-Ni low alloy steel, which is a commercial RPV material. The fracture toughness tests were conducted by 3-point bending with pre-cracked charpy (PCVN) specimens according to the ASTM E1921-09c standard method. The temperature dependency of the fracture toughness was steeper than those predicted by the standard master curve, while the bainitic SA508 Gr.3 steel fitted well with the standard prediction. In order to properly evaluate the fracture toughness of the Gr.4N steels, the exponential coefficient of the master curve equation was changed and the modified curve was applied to the fracture toughness test results of model alloys that have various chemical compositions. It was found that the modified curve provided a better description for the overall fracture toughness behavior and adequate T{sub 0} determination for the tempered martensitic SA508 Gr.4N steels.

  17. Study of atomic clusters in neutron irradiated reactor pressure vessel surveillance samples by extended X-ray absorption fine structure spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)], E-mail: Sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Luetzenkirchen-Hecht, D.; Frahm, R. [Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)

    2009-03-31

    Copper and nickel impurities in nuclear reactor pressure vessel (RPV) steel can form nano-clusters, which have a strong impact on the ductile-brittle transition temperature of the material. Thus, for control purposes and simulation of long irradiation times, surveillance samples are submitted to enhanced neutron irradiation. In this work, surveillance samples from a Swiss nuclear power plant were investigated by extended X-ray absorption fine structure spectroscopy (EXAFS). The density of Cu and Ni atoms determined in the first and second shells around the absorber is affected by the irradiation and temperature. The comparison of the EXAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the EXAFS analysis reveals local irradiation damage in the form of vacancy fractions, which can be determined with a precision of {approx}5%. There are indications that the formation of Cu and Ni clusters differs significantly.

  18. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  19. Study on probability of failure for RPV nozzle region under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Il Soon; Jeong, Kwang Jin; Oh, Young Jin; Kwon, Seung Uk; Jun, Hyun Chul [Seoul National Univ., Seoul (Korea)

    2001-04-01

    Most of previous study for creep rupture of RPV lower head under severe accident condition, have been focused on global failure of RPV lower head. In contract, the local failure of the RPV nozzle region has not been studied in detail. The existence and features of nozzle failure in LAVA-ICI specimen of KAERI and LHF-4 specimen of Sandia National Lab., are observed. It is confirmed that the nozzle failure of LHF-4 specimen is due to the hoop stress in the RPV. The tensile tests in various temperatures and the creep rupture tests in various temperatures and stresses, are accomplished. The finite element analysis for LAVA-ICI experiment was confirmed, and the stress and deformation analysis results are used in LAVA-ICI experiment. 17 refs., 34 figs., 3 tabs. (Author)

  20. The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for 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); Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    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. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

  1. Characterization of transition behavior in SA508 Gr.4N Ni-Cr-Mo low alloy steels with microstructural alteration by Ni and Cr contents

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki-Hyoung; Park, Sang-gyu [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of); Nuclear Materials Research Division, KAERI, Daejeon 305-353 (Korea, Republic of); Kim, Min-Chul, E-mail: mckim@kaeri.re.kr [Nuclear Materials Research Division, KAERI, Daejeon 305-353 (Korea, Republic of); Lee, Bong-Sang [Nuclear Materials Research Division, KAERI, Daejeon 305-353 (Korea, Republic of); Wee, Dang-Moon [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of)

    2011-11-25

    Highlights: {yields} We offer information for determining optimum alloying contents of SA508 Gr.4N steel. {yields} This study shows improvement of toughness with increasing Ni and Cr contents. {yields} Ni content is more effective on the impact toughness than on the fracture toughness. {yields} Cr content is more effective on the fracture toughness. {yields} We offer detailed information on relationship between toughness and microstructure. - Abstract: SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial reactor pressure vessel (RPV) steels, may be a candidate RPV material with the improved strength and toughness due to its tempered martensitic microstructure. This study aims at assessing the effects of microstructural factors with alloying element contents on the transition properties of Ni-Cr-Mo low alloy steels. Model alloys with different Ni and Cr contents were fabricated and their Charpy impact toughness and fracture toughness were examined in the transition region according to ASTM E23 and E1921 standard procedures, respectively. The test results showed extensive improvements of both impact toughness and fracture toughness with increasing Ni and Cr contents. However, Ni content was more effective on the impact toughness than on the fracture toughness, while Cr content was more effective on the fracture toughness. In order to identify a difference in effects of alloying elements contents on the fracture toughness and impact toughness, the relations between the transition properties and the scale of the microstructural features such as packets and carbides are discussed in detail.

  2. Finite element analysis to estimate burst pressure of mild steel pressure vessel using Ramberg–Osgood model

    OpenAIRE

    Deolia, Puneet; Firoz A. Shaikh

    2016-01-01

    Burst pressure is the pressure at which vessel burst/crack and internal fluid leaks. An accurate prediction of burst pressure is necessary in chemical, medical and aviation industry. Burst pressure is a design safety limit, which should not be exceeded. If this pressure is exceeded it may lead to the mechanical breach and permanent loss of pressure containment. So burst pressure calculation is necessary for all the critical applications. To numerically calculate burst pressure material curve ...

  3. Results of work in the hot cells of Laboratory Testing Materials Irradiated Areva of Carina project for the expansion of the database of mechanical characteristics of fractures in materials of RPV German irradiated; Resultados del trabajo en las celdas calientes del Laboratorio de Ensayos de Materiales Irradiados de Areva del proyecto Carina para la ampliacion de la base de datos de caracteristicas mecanicas de las fracturas en materiales de RPV alemanas irradiados

    Energy Technology Data Exchange (ETDEWEB)

    Barthelmes, J.; Schabel, H.; Hein, H.; Kein, E.; Eiselt, C.

    2013-07-01

    In the frame of the already completed research projects CARINA and its predecessor CARISMA a data base was created for pre-irradiated original RPV steels of German PWRs which allowed to examine the consequences if the Master Curve (T{sub 0}) approach instead of the RT{sub N}OT concept is applied to the RPV safety assessment. Furthermore in CARINA different irradiation conditions with respect to the accumulated neutron fluences and specific impact parameters were investigated. Besides a brief introduction of the CARINA project and an overview of the main results an overview on the requirements of the hot laboratory work in terms of specimen manufacturing and material testing is given and examples for realization are shown. (Author)

  4. A study on material degradation in SB 410 carbon steel plates for boilers and other pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Baek, U. B.; Park, J. S. [Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of); Nam, K. W.; Kim, H. Y. [Korea Energy Management Corp., Yongin (Korea, Republic of)

    2005-07-01

    In spite of frequent defect in industrial boilers, life assessment or diagnostic method for them has not been studied. In this research, SB410 carbon steel used in industrial boilers is simulated with artificial aging heat treatment. To do qualitative life assessment, differences in micro-structures and hardness of SB410 by the degradation time are studied. In addition, variation in material properties by aging was observed with the tensile test at room temperature and 179 .deg. C and changes in ductile to brittle transition temperature was observed with the charpy impact test performed at several test temperature.

  5. Method for LEFM analysis of RPV during SBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.

    1985-01-01

    A somewhat simplified two-dimensional linear elastic fracture mechanics model of the beltline of a RPV is presented. The fracture mechanics model used tends to be conservative in the sense that it ignores possible beneficial effects of warm prestressing and cladding. For LEFM studies that require a large number of analyses on the same geometry but with different loads and material toughnesses, the superposition principle is an accurate and simple method to determine K/sub 1/, provided that K/sub 1/ due to a unit load (called K* in this paper), acting on an arbitrary point on the crack surface is known. The details of the superposition principle and the procedure used for determining K* have been presented. It is believed that the error in the K/sub 1/ - values so determined is less than 3.5%. Once these K* - are determined for a specific geometry, then the determination of K/sub 1/ for the same geometry can be made accurately and in a manner that permits parametric studies involving thousands of individual analyses. An example of the use of the simplified model for a parametric analysis is also presented. 35 refs., 11 figs., 1 tab.

  6. Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2001-03-01

    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

  7. 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.

  8. Qualification of phased array ultrasonic examination on T-joint weld of austenitic stainless steel for ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G.H. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Park, C.K., E-mail: love879@hanmail.net [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Jin, S.W.; Kim, H.S.; Hong, K.H.; Lee, Y.J.; Ahn, H.J.; Chung, W. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Jung, Y.H.; Roh, B.R. [Hyundai Heavy Industries Co. Ltd., Ulsan 682-792 (Korea, Republic of); Sa, J.W.; Choi, C.H. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • PAUT techniques has been developed by Hyundai Heavy Industries Co., LTD (HHI) and Korea Domestic Agency (KODA) to verify and settle down instrument calibration, test procedures, image processing, and so on. As the first step of development for PAUT technique, Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. Real UT qualification group-1 for T-joint weld was successfully conducted in front of ANB inspector. • In this paper, remarkable progresses of UT qualification are presented for ITER vacuum vessel. - Abstract: Full penetration welding and 100% volumetric examination are required for all welds of pressure retaining parts of the ITER Vacuum Vessel (VV) according to RCC-MR Code and French Order of Nuclear Pressure Equipment (ESPN). The NDE requirement is one of important technical issues because radiographic examination (RT) is not applicable to many welding joints. Therefore the ultrasonic examination (UT) has been selected as an alternative method. Generally the UT on the austenitic welds is regarded as a great challenge due to the high attenuation and dispersion of the ultrasonic signal. In this paper, Phased array ultrasonic examination (PAUT) has been introduced on double sided T-shape austenitic welds of the ITER VV as a major NDE method as well as RT. Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, embedded parallel vertical notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. PAUT techniques on the thick austenitic welds have been developed taking into account the acceptance criteria. Test procedure including calibration of equipment is derived and qualified through

  9. Intergranular fracture stress and phosphorus grain boundary segregation of a Mn-Ni-Mo steel

    Energy Technology Data Exchange (ETDEWEB)

    Naudin, C.; Frund, J.M. [EDF, Moret sur Loing (France). Direction des Etudes et Recherches; Pineau, A. [Ecole des Mines de Paris, Evry (France). Centre des Materiaux

    1999-04-09

    Nuclear Reactor Pressure Vessel (RPV) steel A508 class 3 which is a low alloyed steel is not usually sensitive to reversible temper embrittlement when properly heat treated. However heterogeneous zones may be present in particular near the inner side of the vessel. These zones result from the segregation of the alloying elements (C, Mn, Ni, Mo) and impurities (S, P) taking place during solidification of the material. They are called segregated zones (or ghost lines). They can reach 2 mm thick along the radius and 30 mm long through the circumferential direction. Their susceptibility to reversible temper embrittlement is mainly due to grain boundary phosphorus segregation triggering brittle intergranular fracture when the material is tested at low temperature. In this material like in other steels the influence of some other alloying elements (Mo, Mn...) is clearly significant and should also be taken into account. But phosphorus effect has proved to be predominant. The aim of the present study is therefore to find out a quantitative relationship between grain boundary phosphorus segregation and critical intergranular fracture stress. A synthetic steel with a chemical composition representative of an average segregated zone was prepared for the present study. A number of heat treatments were applied to reach different embrittlement conditions. Then brittle fracture properties were obtained by performing cryogenic fracture tests on notched tensile specimens while the corresponding grain boundary phosphorus levels were measured by Auger electron spectroscopy. Systematic fractographic observations were carried out. Moreover an attempt to determine the influence of temperature on the critical intergranular fracture stress was made.

  10. Evaluation of Nozzle Arrangement Focused on RPV Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Wook; Lee, Gyu Mahn; Jeoung, Kyeong Hoon; Kim, Tae Wan; Park, Keun Bae; Kim, Keung Koo

    2008-10-15

    The purpose of this study is to investigate the fabrication capacity of the reactor pressure vessel. For that reason, this study focuses on survey of the domestic equipment capacity and the feasible size for reactor pressure vessel. Also, the forecasting issues of adoption of new material for reactor pressure vessel are reviewed through typically examples. Additionally, an evaluation procedure for the design of nozzle is developed to meet ASME code requirements. The developed design procedure could provide typical references for the development of advanced reactor design in the future.

  11. Finite element analysis to estimate burst pressure of mild steel pressure vessel using Ramberg–Osgood model

    Directory of Open Access Journals (Sweden)

    Puneet Deolia

    2016-09-01

    Full Text Available Burst pressure is the pressure at which vessel burst/crack and internal fluid leaks. An accurate prediction of burst pressure is necessary in chemical, medical and aviation industry. Burst pressure is a design safety limit, which should not be exceeded. If this pressure is exceeded it may lead to the mechanical breach and permanent loss of pressure containment. So burst pressure calculation is necessary for all the critical applications. To numerically calculate burst pressure material curve is essential. There are various material models which are used to define material curve, amongst them Ramberg–Osgood is very popular. Ramberg–Osgood accurately capture material curve in strain hardening region. This approach is applicable for different material grades. In this paper a finite element method is used to predict burst pressure using Ramberg–Osgood equation. These results are then compared with results obtained from elasto-plastic curve and true stress strain curve. Results obtained by finite element analysis are validated with experimental data which is considered from open literature.

  12. Analysis of the master curve approach on the fracture toughness properties of SA508 Gr.4N Ni-Mo-Cr low alloy steels for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki-Hyoung, E-mail: shirimp@kaist.ac.kr [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of); Kim, Min-Chul; Lee, Bong-Sang [Nuclear Materials Research Division, KAERI, Daejeon 305-353 (Korea, Republic of); Wee, Dang-Moon [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of)

    2010-06-15

    This study aims at assessing the fracture toughness behavior of tempered martensitic Ni-Mo-Cr low alloy steels for reactor pressure vessels in a transition temperature region using a master curve approach. The fracture toughness tests for model alloys with various chemical compositions were carried out following ASTM E1921-08. The microstructures, tensile properties, and Charpy impact toughness were also evaluated. Alloying elements such as Ni, Cr, and Mo affected the mechanical properties of alloys from changes in the phase fraction and precipitation behavior. In the fracture toughness test results, the data sets showed a deviation from the median curve and a smaller scatter than that of the prediction of the ASTM standard, especially in the lower transition region. The exponential parameter of the master curve equation was adjusted by an exponential fitting to data sets for expressing well the temperature dependency of the fracture toughness. The adjusted parameter provided good agreement for data distribution and the independence of T{sub 0} from test temperatures through an overall temperature range in contrast with the results from the standard master curve.

  13. Annealing for plant life management: hardness, tensile and Charpy toughness properties of irradiated, annealed and re-irradiated mock-up low alloy nuclear pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Tipping, Philip; Cripps, Robin (Paul Scherrer Inst. (PSI), Villigen (Switzerland))

    1994-01-01

    Hardness, tensile and Charpy properties of an irradiated (I) and irradiated-annealed-reirradiated (IAR) mock-up pressure vessel steel are presented. Spectrum tailored pressurized light water reactor (PWR) irradiation at 290[sup o]C by fast neutrons up to nominal fluences of 5 x 10[sup 19]/cm[sup 2] (E [>=] 1 MeV) in a swimming pool type reactor caused the hardness, tensile yield stress and tensile strength to increase. Embrittlement also occurred as indicated by Charpy toughness tests. The optimum annealing heat treatment for the main program was determined using isochronal and isothermal runs on the material and measuring the Vickers microhardness. The response to an intermediate annealing treatment (460[sup o]C for 18 h), when 50% of the target fluence has been reached and then irradiating to the required end fluence (IAR condition) was then monitored further by Charpy and tensile mechanical properties. Annealing was beneficial in mitigating overall hardening or embrittlement effects. The rate of re-embrittlement after annealing and re-irradiating was no faster than when no annealing had been performed. Annealing temperatures below 440[sup o]C were indicated as requiring relatively long times, i.e. [>=] 168 h to achieve some reduction in radiation induced hardness for example. (Author).

  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. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  16. 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

  17. Influence of operation factors on brittle fracture initiation and critical local normal stress in SE(B) type specimens of VVER reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Erak, A. D.; Kiselev, A. S.; Bubyakin, S. A.; Bandura, A. P.

    2015-12-01

    A complex of mechanical tests and fractographic studies of VVER-1000 RPV SE(B) type surveillance specimens was carried out: the brittle fracture origins were revealed (non-metallic inclusions and structural boundaries) and the correlation between fracture toughness parameters (CTOD) and fracture surface parameters (CID) was established. A computational and experimental method of the critical local normal stress determination for different origin types was developed. The values of the critical local normal stress for the structural boundary origin type both for base and weld metal after thermal exposure and neutron irradiation are lower than that for initial state due to the lower cohesive strength of grain boundaries as a result of phosphorus segregation.

  18. SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E.; Chaouadi, R.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J

    1998-05-04

    The contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV weld material is reported. The objective of this contribution is twofold: (1) to gain experience in the field of the testing of WWER-440 steels; (2) to analyse the round-robin data according to in-house developed on used models in order to check their validity and applicability. Results from testing on unirradiated material are reported including data obtained from chemical analysis, Charpy-V impact testing, tensile testing and fracture toughness determination. Finally, irradiation strategies that can be used in the program to obtain irradiated, irradiated-annealed and irradiated-annealed-reirradiated conditions are outlined.

  19. Searches for RPV SUSY and long-lived particles at the LHC

    CERN Document Server

    Liu, Minghui; The ATLAS collaboration

    2015-01-01

    Both the ATLAS and CMS collaboration have made great effort to search for RPV SUSY and LLP. Tens of models are used to perform studies, and all the observations seem to be in good agreement with background expectation. Most stringent limits up to date are put on these new models.

  20. Continuous steel production and apparatus

    Science.gov (United States)

    Peaslee, Kent D [Rolla, MO; Peter, Jorg J [McMinnville, OR; Robertson, David G. C. [Rolla, MO; Thomas, Brian G [Champaign, IL; Zhang, Lifeng [Trondheim, NO

    2009-11-17

    A process for continuous refining of steel via multiple distinct reaction vessels for melting, oxidation, reduction, and refining for delivery of steel continuously to, for example, a tundish of a continuous caster system, and associated apparatus.

  1. Development of automatic reactor vessel inspection systems; development of data acquisition and analysis system for the nuclear vessel weld

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Po; Park, C. H.; Kim, H. T.; Noh, H. C.; Lee, J. M.; Kim, C. K.; Um, B. G. [Research Institute of KAITEC, Seoul (Korea)

    2002-03-01

    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine heavy vessel welds. In order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet. In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed. In this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition software was developed. The new systems were tested on the RPV welds of Ulchin Unit 6 to confirm their functions and capabilities. They worked very well as designed and the tests were successfully completed. 13 refs., 34 figs., 11 tabs. (Author)

  2. Investigation for deformation of ion-irradiated RPV steel using nanoindentation hardness test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon; Kim, In Sub [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-10-01

    To evaluate deformation depending on depth under ion irradiation, nanoindentation test was applied to heavy ion-irradiated SA508 CL.3 RPVSs. The specimens were irradiated with 8MeV Fe{sup +4} ions to 0.15dpa and 1.5dpa at below 60 .deg. C. The derivative of the load-depth ratio, d(L/D)/dD, was estimated the depth dependent formation of plastic and elastic deformation in the irradiated specimens. In ion-irradiated SA508 CL.3 RPVS, the peak of deformation was observed at about 20nm from the surface, but the one of radiation damage was appeared at 1500nm from the surface when TRIM98 was simulated damage depth profile. In order to study the effect of deformation depth under ion irradiation rate, we evaluated nanoindentation hardness test. The hardness was larger for the irradiated than for the non-irradiated, and also larger for 1.5dpa than for 0.15dpa one.

  3. Historical introgression of the downy mildew resistance gene Rpv12 from the Asian species Vitis amurensis into grapevine varieties.

    Directory of Open Access Journals (Sweden)

    Silvia Venuti

    Full Text Available The Amur grape (Vitis amurensis Rupr. thrives naturally in cool climates of Northeast Asia. Resistance against the introduced pathogen Plasmopara viticola is common among wild ecotypes that were propagated from Manchuria into Chinese vineyards or collected by Soviet botanists in Siberia, and used for the introgression of resistance into wine grapes (Vitis vinifera L.. A QTL analysis revealed a dominant gene Rpv12 that explained 79% of the phenotypic variance for downy mildew resistance and was inherited independently of other resistance genes. A Mendelian component of resistance-a hypersensitive response in leaves challenged with P. viticola-was mapped in an interval of 0.2 cM containing an array of coiled-coil NB-LRR genes on chromosome 14. We sequenced 10-kb genic regions in the Rpv12(+ haplotype and identified polymorphisms in 12 varieties of V. vinifera using next-generation sequencing. The combination of two SNPs in single-copy genes flanking the NB-LRR cluster distinguished the resistant haplotype from all others found in 200 accessions of V. vinifera, V. amurensis, and V. amurensis x V. vinifera crosses. The Rpv12(+ haplotype is shared by 15 varieties, the most ancestral of which are the century-old 'Zarja severa' and 'Michurinets'. Before this knowledge, the chromosome segment around Rpv12(+ became introgressed, shortened, and pyramided with another downy mildew resistance gene from North American grapevines (Rpv3 only by phenotypic selection. Rpv12(+ has an additive effect with Rpv3(+ to protect vines against natural infections, and confers foliar resistance to strains that are virulent on Rpv3(+ plants.

  4. 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.

  5. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  6. Unmanned Air Vehicle/Remotely Piloted Vehicle Analysis for Lethal UAV/ RPV

    Science.gov (United States)

    1993-09-01

    the fuselage or one or two piece modular design with glass fiber skin and rigid p:)ly-vinyl chloride injected foam core, wood frame with veneer skin, or...materials such as carbon fiber, kevlar, and epoxy resins used in RPV structural design. In fact, almost all RPVs composites. Structural designs utilize...fiberglass, honeycomb, molded glass fiber-reinforced plastics (GRP) or wound glass fiber impregnated with resin . Wings may be molded integrally with

  7. The Design of a FLIR Sensor for the Korean Army RPV

    Science.gov (United States)

    1991-09-01

    IR system include properties of terrestrial materials of the operation area and atmospheric turbulence. Terrestrial properties involve soil , rocks...BOTTOM VIEW) . 11 FIGURE 2.4 RPV FLIGHT SCENARIO ...................... 10 FIGURE 4.1 COMMON MODULE FLIR ZERO AZIMUTH SCAN . 23 FIGURE 4.2 NARCISSUS OF...and other agricultural products. In particular, the middle part of the Korean peninsula consists of two specific zones. Soft soils and road construction

  8. Vessel Operating Units (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

  9. Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994

    Energy Technology Data Exchange (ETDEWEB)

    Lowendahl, B. [OKG Aktiebolag, Figeholm (Sweden)

    1995-03-01

    In September 1992 Oskarshamn 1 was shut down in order to carry out measures to correct discovered deficiencies in the emergency cooling systems. Due to the results of a comprehensive non destructive test programme it was decided to perform a major replacement of pipes in the primary systems including a full system decontamination using the Siemens CORD process. The paper briefly presents the satisfying result of the decontamination performed in May-June 1993. When in late June 1993 cracks also were detected in the feed-water pipes situated inside the reactor pressure vessel (RPV) the plans were reconsidered and a large project was formed with the aim, in a first phase, to verify the integrity of the RPV. In order to make it possible to perform work manually inside the RPV special radiation protection measures had to be carried out. In January 1994 the lower region of the RPV was decontaminated, again using the CORD-process, followed by the installation of a special shielding construction in the RPV. The surprisingly good results of these efforts are also briefly described in the paper.

  10. Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

    Energy Technology Data Exchange (ETDEWEB)

    Williams, P. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dickson, T. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yin, S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2007-12-01

    The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  12. 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.

  13. Development of Mono-bloc Forging for CAP1400 Reactor Pressure Vessel

    Science.gov (United States)

    Bao-zhong, Wang; Kai-quan, Liu; Ying, Liu; Wen-hui, Zhang; De-li, Zhao

    With the Development of Larger Output and Longer Plant Life for Advanced Pressurized Water Reactor Pressure Vessel (APWRPV), Larger and Mono-bloc Forgings must be Necessary. In the researching and manufacturing of CAP1400RPV forgings, China First Heavy Industries (CFHI) Manufactured Mono-bloc Upper Head with Quick-loc and Mono-bloc Lower Head (Combined Part of Transition Ring and Lower Head),reducing the welding process and the mechanical property deference comparing with the traditional manufacturing, Comparing with AP1000 forgings, not only Obtained better Mechanical Properties, but also Reduced Manufacturing cycle of RPV and Nondestructive in-service Inspection due to Elimination of Weld Seams. CHFI will Development Mono-bloc Nozzle Shell (Combined Part of Vessel Flange Nozzle Shell and Inlet/Outlet Nozzles) for Future APWRPV.

  14. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Mizutani, J.; Kawamura, S. [Tokyo Electric Power Co., Inc. (Japan); Aoki, M. [Hitachi Ltd., Ibaraki (Japan); Mori, T. [Toshiba Corp., Yokihama (Japan)

    2001-07-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  15. Overview of feasibility study on conducting overflight measurements of shaped sonic boom signatures using RPV's

    Science.gov (United States)

    Maglieri, Domenic J.; Sothcott, Victor E.; Keefer, Thomas N., Jr.; Bobbitt, Percy J.

    1992-04-01

    Before beginning this presentation, it is appropriate to acknowledge the sincere interest and financial support provided by the NASA LaRC under contract NAS9-17900. An outline of the material to be used in the present paper is given. It begins with a indication of the origin and objectives of the feasibility study. This is followed by a discussion of various simulation methods of establishing the persistence of shaped sonic boom signatures to large distances including the use of recoverable RPV/drones. The desirable features to be sought out in an RPV along with a rationale for the selection of a 'shaped' sonic boom signature will be addressed. Three candidate vehicles are examined as to their suitability with respect to a number of factors, in particular, modifiability. Area distributions and associated sonic boom signatures of the basic and modified Firebee vehicle will also be shown. An indication of the scope of the proposed wind tunnel and flight test programs will be presented including measurement technologies and predicted waveforms. Finally, some remarks will be made summarizing the study and highlighting the key findings.

  16. 46 CFR 59.20-1 - Carbon-steel or alloy-steel castings.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Carbon-steel or alloy-steel castings. 59.20-1 Section 59... BOILERS, PRESSURE VESSELS AND APPURTENANCES Welding Repairs to Castings § 59.20-1 Carbon-steel or alloy-steel castings. Defects in carbon-steel or alloy-steel castings may be repaired by welding. The repairs...

  17. Application of Remotely Piloted Vehicle (RPV) in monitoring and detecting watershed land use change and problem areas

    Science.gov (United States)

    Long-Ming Huang

    2000-01-01

    Improper cultivation of steep mountainous areas in Taiwan contributes to serious erosion and landslides. Regular patrol, detection, and administration of these problem areas has been an extremely difficult due to the steep and dangerous terrain of many of the forested watersheds in Taiwan. A remotely piloted vehicle (RPV) has been developed for various civil and...

  18. Effects of tempering and PWHT on microstructures and mechanical properties of SA508 Gr.4N steel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Jhung, Myung Jo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-06-15

    Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT) conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV) material. The blocks of model alloy were austenitized at the conventional temperature of 880 degrees Celsius then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is significantly improved, especially in the specimens tempered at 630 degrees Celsius. The sample tempered at 630 degrees Celsius with PWHT at 610 degrees Celsius shows optimum mechanical properties in hardness, strength, and toughness, excluding only the transition property in the low temperature region. The microstructural observation and quantitative analysis of carbide size distribution show that the variations of mechanical properties are caused by the under-tempering and carbide coarsening which occurred during the heat treatment process. The introduction of PWHT results in the deterioration of the ductile-brittle transition property by an increase of coarse carbides controlling cleavage initiation, especially in the tempered state at 630 degrees Celsius.

  19. EFFECTS OF TEMPERING AND PWHT ON MICROSTRUCTURES AND MECHANICAL PROPERTIES OF SA508 GR.4N STEEL

    Directory of Open Access Journals (Sweden)

    KI-HYOUNG LEE

    2014-06-01

    Full Text Available Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV material. The blocks of model alloy were austenitized at the conventional temperature of 880 °C, then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is significantly improved, especially in the specimens tempered at 630 °C. The sample tempered at 630 °C with PWHT at 610 °C shows optimum mechanical properties in hardness, strength, and toughness, excluding only the transition property in the low temperature region. The microstructural observation and quantitative analysis of carbide size distribution show that the variations of mechanical properties are caused by the under-tempering and carbide coarsening which occurred during the heat treatment process. The introduction of PWHT results in the deterioration of the ductile-brittle transition property by an increase of coarse carbides controlling cleavage initiation, especially in the tempered state at 630 °C.

  20. Strain-based plastic instability acceptance criteria for ferritic steel safety class 1 nuclear components under level D service loads

    Directory of Open Access Journals (Sweden)

    Ji-Su Kim

    2015-04-01

    Full Text Available This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1 a section average approach and (2 a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the overconservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

  1. Switching from an EFV-based STR to a RPV-based STR is effective, safe and improves HIV patients health status.

    Science.gov (United States)

    Maggiolo, Franco; Di Matteo, Sergio; Bruno, Giacomo; Astuti, Noemi; Di Filippo, Elisa; Valenti, Daniela; Colombo, Giorgio

    2014-01-01

    TDF/FTC/RPV has been shown effective in both naïve and PI-pre-treated patients. Less is known about a switch strategy in subjects receiving EFV. We evaluated viro-immunologic outcomes, Quality of Life (QoL) and costs of an unselected cohort of patients switching from a TDF/FTC/EFV STR (≥6 months duration) to a TDF/FTC/RPV STR. The considered outcome measures were quality-adjusted life years (QALYs) as measured with the EQ5D questionnaire and the overall direct health costs. 64 patients with a baseline viral loadEFV for four months and then switch to TDF/FTC/RPV. Six patients in the deferred switch group did not actually change cART. Patients were mostly males (73.4%) with a mean age of 46 years, a baseline mean HIV-RNA of 6.4 copies/mL and a mean baseline CD4 count of 588 cells/µL. For the considered follow-up period, the mean cost per patient resulted 2,563 for TDF/FTC/RPV and 2,572 for TDF/FTC/EFV. Viremia remained undetectable and CD4 stable in all patients. Over time the mean QoL increased in the RPV arm ad slightly decreased in the EFV arm, after four months the mean per patient QALYs was 0.849 for RPV and 0.841 for EFV, respectively (Figure 1). A sharp increment of QoL was observed in the deferred-switch arm after switch, too. VAS analysis of health status perception also increased overall from 82.78 to 83.79 due to the improvement in the RPV arm. Mean cholesterol levels improved in the RPV arm from 203 to 170 mg/dL, while an increment from 190 to 207 mg/dL was observed in the EFV arm. HDL levels lowered from 49 to 45 and rose from 53 to 54 mg/dL in the RPV and EFV arms, respectively. Triglycerides levels improved both in the RPV arm (from 138 to 112 mg/dL) and in the EFV arm (from 110 to 103 mg/dL). Switching from TDF/FTC/EFV to TDF/FTC/RPV is a safe, well tolerated strategy that improves the overall health status of HIV-treated patients. The switch does not expose patients to a risk of virologic failure due to possible PK interactions of the drugs. RPV

  2. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    Directory of Open Access Journals (Sweden)

    Weimin Ma

    2016-03-01

    Full Text Available A historical review of in-vessel melt retention (IVR is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs. The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

  3. Current understanding of the effects of enviromental and irradiation variables on RPV embrittlement

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.; Wirth, B.; Liu, C.L. [Univ. of California, Santa Barbara, CA (United States)

    1997-02-01

    Radiation enhanced diffusion at RPV operating temperatures around 290{degrees}C leads to the formation of various ultrafine scale hardening phases, including copper-rich and copper-catalyzed manganese-nickel rich precipitates. In addition, defect cluster or cluster-solute complexes, manifesting a range of thermal stability, develop under irradiation. These features contribute directly to hardening which in turn is related to embrittlement, manifested as shifts in Charpy V-notch transition temperature. Models based on the thermodynamics, kinetics and micromechanics of the embrittlement processes have been developed; these are broadly consistent with experiment and rationalize the highly synergistic effects of most important irradiation (temperature, flux, fluence) and metallurgical (copper, nickel, manganese, phosphorous and heat treatment) variables on both irradiation hardening and recovery during post-irradiation annealing. A number of open questions remain which can be addressed with a hierarchy of new theoretical and experimental tools.

  4. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Tricot, N. [Institut de Radioprotection et de Surete Nucleaire, IRSN/DES/SECCA, 92 - Fontenay aux Roses (France); Jendrich, U. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  5. Initial Evaluation of the Heat-Affected Zone, Local Embrittlement Phenomenon as it Applies to Nuclear Reactor Vessels

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, D.E.

    1999-09-01

    The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds.

  6. Uranyl-Fluoride (235U) Solutions in Spherical Stainless Steel Vessels with Reflectors of Be, Ch2 and Be-Ch2 Composites, Part II

    Energy Technology Data Exchange (ETDEWEB)

    Heinrichs, D

    2002-04-08

    A series of criticality studies were performed at the Lawrence Livermore National Laboratory in the late 1950's using aqueous solutions of {sup 233}U in the form of UO{sub 2}F{sub 2} stabilized with 0.3% by weight of HF. These experiments were assigned the program name Falstaff. The {sup 233}U concentration in these experiments ranged from 0.13 to 0.87 kg/l. Eight type 347 stainless steel spheres ranging in inner radius from 7.87 to 12.45 cm were available for use as containers for the solutions. The scope of this evaluation is limited to the experiments involving the four lowest concentrations of uranyl-fluoride solution with 0.45, 0.37, 0.24 and 0.13 kg ({sup 233}U)/l. Reflectors of beryllium, polyethylene and beryllium-polyethylene composites were used. Thirty-one configurations are evaluated and accepted as criticality-safety benchmark models. Fission rate data calculated by the evaluator (see Appendix B) show that twenty-six of these configurations have over 50% of the fissions occurring in the thermal energy range and these configurations are therefore classified as ''THERMAL''. Five of the configurations have less than 50% of the fissions occurring in any of the fast, intermediate or thermal energy range and therefore are classified as ''MIXED''.

  7. Integral Steel Casting of Full Spade Rudder Trunk Carrier Housing for Supersized Container Vessels through Casting Process Engineering (Sekjin E&T

    Directory of Open Access Journals (Sweden)

    Tae Won Kim

    2015-04-01

    Full Text Available In casting steel for offshore construction, integral casted structures are superior to welded structures in terms of preventing fatigue cracks in the stress raisers. In this study, mold design and casting analysis were conducted for integral carrier housing. Casting simulation was used for predicting molten metal flow and solidification during carrier housing casting, as well as the hot spots and porosity of the designed runner, risers, riser laggings, and the chiller. These predictions were used for deriving the final carrier housing casting plan, and a prototype was fabricated accordingly. A chemical composition analysis was conducted using a specimen sampled from a section of the prototype; the analytically obtained chemical composition agreed with the chemical composition of the existing carrier housing. Tensile and Charpy impact tests were conducted for determining the mechanical material properties. Carrier housing product after normalizing (920 °C/4.5 h, air-cooling has 371 MPa of yield strength, 582 MPa of tensile strength, 33.4% of elongation as well as 64 J (0 °C of impact energy.

  8. Effect of the damage by radiation on the reference temperature T{sub 0} of ferritic steel; Efecto del dano por radiacion en la temperatura de referencia T{sub 0} de acero ferritico

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva O, A

    2004-07-01

    Presently work studies the effect that produces the irradiation in ferritic steels, on the reference temperature T{sub 0} (intrinsic characteristic of the fracture tenacity in the area of ductile-fragile transition), applying the approach of the Master curve that is based on the norm Astm E-1921. For it it was elaborated a methodology and procedure for test tubes type Charpy according to the standard before mentioned. Due to the ferritic steels are used mainly in pressure vessels to the reactor (RPV) of nuclear power plants; in the samples it was simulated the effect of the damage for irradiation through a thermal treatment that induced the precipitation of the carbides and sulfurs in the limits of grain (one of the modifications suffered in the irradiated materials); it was made a comparison later with material samples in initial state (without thermal treatment), used as witness sample, by means of assays of fracture mechanics, specifically flexion in three points; this way with it to observe the effect of the damage for irradiation in the reference temperature (T{sub 0}). This temperature (T{sub 0}) it is a very important parameter in the mechanical property of the material called fracture tenacity; which at the moment gives the rule for the verification of structural integrity of the RPV. As a result of this it was observed an increase in the reference temperature in the material in fragilezed state with respect to the initial state of 31.75 C. They were carried out metallographic analysis and fractographs of the assayed surface finding carbide inclusions and sulfurs that in theory of the Master Curve they are initiators of cracks and of a possible catastrophic flaw of the material. At the moment the Division of Scientific Investigation of the ININ is carrying out activities in the Nucleo electric Central of Laguna Verde (CNLV) related with the program of surveillance of the materials of the vessel of the unit 2, as well as projects of structural integrity

  9. Technical aspects of the process of segmentation and packaging of the reactor vessel of Jose Cabrera NPP; Aspectos tecnicos del proceso de segmentacion y embalaje de la vasija del reactor de la central nuclear Jose Cabrera

    Energy Technology Data Exchange (ETDEWEB)

    Valdivieso, J. M.; Garcia Castro, R.

    2015-07-01

    Westinghouse is carrying out the segmentation of the reactor pressure vessel (RPV) within the framework of the Dismantling and Decommissioning Project of the Jose Cabrera NPP. The final concept is based on the comprehensive Westinghouse experience in the field of LWR pressure vessel and internals segmentation, and particularly in previous reactor internals segmentation project for Jose Cabrera NPP. This article shows the development of all the activities included: cutting method selection, preparatory works, cutting activities, waste characterization and packaging activities. (Author)

  10. First principle-based AKMC modelling of the formation and medium-term evolution of point defect and solute-rich clusters in a neutron irradiated complex Fe–CuMnNiSiP alloy representative of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Ngayam-Happy, R., E-mail: raoul.ngayamhappy@gmail.com [EDF-R and D, Département Matériaux et Mécanique des Composants (MMC), Les Renardières, F-77818 Moret sur Loing Cedex (France); Unité Matériaux et Transformations (UMET), UMR CNRS 8207, Université de Lille 1, ENSCL, F-59655 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Becquart, C.S. [Unité Matériaux et Transformations (UMET), UMR CNRS 8207, Université de Lille 1, ENSCL, F-59655 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 and D, Département Matériaux et Mécanique des Composants (MMC), Les Renardières, F-77818 Moret sur Loing Cedex (France); Unité Matériaux et Transformations (UMET), UMR CNRS 8207, Université de Lille 1, ENSCL, F-59655 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France)

    2013-09-15

    The formation and medium-term evolution of point defect and solute-rich clusters under neutron irradiation have been modelled in a complex Fe–CuMnNiSiP alloy representative of RPV steels, by means of first principle-based atomistic kinetic Monte Carlo simulations. The results obtained reproduce most features observed in available experimental studies, highlighting the very good agreement between both series. According to simulation, solute-rich clusters form and develop via an induced segregation mechanism on either the vacancy or interstitial clusters, and these point defect clusters are efficiently generated only in cascade debris and not Frenkel pair flux. The results have revealed the existence of two distinct populations of clusters with different characteristic features. Solute-rich clusters in the first group are bound essentially to interstitial clusters and they are enriched in Mn mostly, but also Ni to a lesser extent. Over the low dose regime, their density increases in the alloy as a result of the accumulation of highly stable interstitial clusters. In the second group, the solute-rich clusters are merged with vacancy clusters, and they contain mostly Cu and Si, but also substantial amount of Mn and Ni. The formation of a sub-population of pure solute clusters has been observed, which results from annihilation of the low stable vacancy clusters on sinks. The results indicate finally that the Mn content in clusters is up to 50%, Cu, Si, and Ni sharing the other half in more or less equivalent amounts. This composition has not demonstrated any noticeable modification with increasing dose over irradiation.

  11. Impact of radiation protection planning on the dismantling strategy of the cylindrical part of the RPV of Wuergassen NPP

    Energy Technology Data Exchange (ETDEWEB)

    Klein, A.; Dittrich, W.; Terry, I. [AREVA NP GmbH, Erlangen (Germany)

    2010-05-15

    The packing of cut segments of the RPV of Wuergassen NPP demands consideration of radiation protection aspects. By an iterative process, which considers on the one hand engineering aspects and on the other hand the specified limits for radiological quantities, an optimized package concept had been compiled which is characterized by a minimum amount of containers. Radiological measurements for the first packed containers are in good agreement with calculated dose rates and comply with the KONRAD disposal conditions. (orig.)

  12. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf (Germany)

    2008-07-01

    WWER-440 second generation (V-213) reactor pressure vessels (RPV) were produced by IZHORA in Russia and by SKODA in the former Czechoslovakia. The surveillance Charpy-V and fracture mechanics SE(B) specimens of both producers have different orientations. The main difference is the crack extension direction which is through the RPV thickness and circumferential for ISHORA and SKODA RPV, respectively. In particular for the investigation of weld metal from multilayer submerged welding seams the crack extension direction is of importance. Depending on the crack extension direction in the specimen there are different welding beads or a uniform structure along the crack front. The specimen orientation becomes more important when the fracture toughness of the weld metal is directly determined on surveillance specimens according to the Master Curve (MC) approach as standardised in the ASTM Standard Test Method E1921. This approach was applied on weld metal of the RPV beltline welding seam of Greifswald Unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The specimens are in TL and TS orientation. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the MC. Nearly all values lie within the fracture toughness curves for 5% and 95% fracture probability. There is a strong variation of the reference temperature T{sub 0} though the thickness of the welding seam, which can be explained with structural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TS and TL orientation in the welding seam have a differentiating and integrating behaviour, respectively. The statistical assumptions behind the MC approach are valid for both specimen orientations even if the structure is not uniform along the crack front. By comparison crack extension, JR, curves measured on SE(B) specimens with TL and TS orientation

  13. Confirmatory investigations on the flux effect and associated unstable matrix damage in RPV materials exposed to high neutron fluence

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R., E-mail: rachid.chaouadi@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Gérard, R. [Tractebel Engineering, Avenue Ariane 7, 1200 Brussels (Belgium)

    2013-06-15

    This paper provides additional experimental data on the neutron flux effect on RPV hardening and embrittlement and on the so-called unstable matrix damage that was suggested to occur at high flux. Six materials taken from the first irradiation surveillance capsules of Belgian PWRs with a fluence not exceeding about 1.5 × 10{sup 19} n/cm{sup 2} were further irradiated in the BR2 high flux reactor to additional fluences ranging between about 1 and 1.5 × 10{sup 20} n/cm{sup 2} at 290 °C. Eight additional RPV materials were selected to investigate the flux effect on irradiation hardening. No statistically-significant difference in irradiation hardening for low and high flux could be evidenced from the null hypothesis test applied with the general linear model. This is confirmed by additional experiments where fourteen irradiated specimens of various RPV materials consisting of low to high Cu and Ni contents were annealed at 350 °C for 5 h to eventually reveal some recovery of the unstable matrix damage. The results did not show any recovery upon heat treatment, which indicates that unstable matrix defects did not appear in these materials during irradiation at high flux.

  14. Allelic variation at the rpv1 locus controls partial resistance to Plum pox virus infection in Arabidopsis thaliana.

    Science.gov (United States)

    Poque, S; Pagny, G; Ouibrahim, L; Chague, A; Eyquard, J-P; Caballero, M; Candresse, T; Caranta, C; Mariette, S; Decroocq, V

    2015-06-25

    Sharka is caused by Plum pox virus (PPV) in stone fruit trees. In orchards, the virus is transmitted by aphids and by grafting. In Arabidopsis, PPV is transferred by mechanical inoculation, by biolistics and by agroinoculation with infectious cDNA clones. Partial resistance to PPV has been observed in the Cvi-1 and Col-0 Arabidopsis accessions and is characterized by a tendency to escape systemic infection. Indeed, only one third of the plants are infected following inoculation, in comparison with the susceptible Ler accession. Genetic analysis showed this partial resistance to be monogenic or digenic depending on the allelic configuration and recessive. It is detected when inoculating mechanically but is overcome when using biolistic or agroinoculation. A genome-wide association analysis was performed using multiparental lines and 147 Arabidopsis accessions. It identified a major genomic region, rpv1. Fine mapping led to the positioning of rpv1 to a 200 kb interval on the long arm of chromosome 1. A candidate gene approach identified the chloroplast phosphoglycerate kinase (cPGK2) as a potential gene underlying the resistance. A virus-induced gene silencing strategy was used to knock-down cPGK2 expression, resulting in drastically reduced PPV accumulation. These results indicate that rpv1 resistance to PPV carried by the Cvi-1 and Col-0 accessions is linked to allelic variations at the Arabidopsis cPGK2 locus, leading to incomplete, compatible interaction with the virus.

  15. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    by the research vessels RV Gaveshani and ORV Sagar Kanya are reported. The work carried out by the three charted ships is also recorded. A short note on cruise plans for the study of ferromanganese nodules is added...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

  17. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  18. Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.

  19. Status and Perspectives of Nuclear Reactor Pressure Vessel Life Extension up to 60 Years Operation in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Chaouadi, R.; Scibetta, M.; Van Walle, E.

    2009-09-15

    The scope of this report involves a safety evaluation of the reactor pressure vessel (RPV) against neutron embrittlement, in the most severely irradiation region (belt line) and in the event of a pressurized thermal shock. The irreplaceable RPV is considered to be the most critical component for lifetime considerations of the nuclear power plant. However, an application for operation extension will also depend upon a number of additional considerations, including the technical assessment of other plant components, as well as non-technical arguments (e.g. political, environmental, economical, strategical that are outside the scope this report. In the hypothesis of a request for operation extension, it is the responsibility of the utilities to provide the safety authorities with an exhaustive dossier demonstrating that safe extended operation is guaranteed. The role of the safety authorities is to critically evaluate the safety dossier for eventually granting the operation extension.

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

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States); Sen, Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically

  1. AQUILA Remotely Piloted Vehicle System Technology Demonstrator (RPV-STD) Program. Volume I. System Description and Capabilities

    Science.gov (United States)

    1979-04-01

    the duct lip just Inside the leading edge. The struts are 0.5 in. thick. Molded slots on the fuselage end of the struts permit positioning on and...STARTER MOTORE R BE6- A RN O - (3N 0I ) (057563) BUMPER (695ŕ) Ow )AJMNTAse ly HAPOTT (FOUR PLACES) Flgwe 130. Starter Motor Aebly 274 t . . • ,, i II The...and is stopped by a rubber bumper assembly. To prevent rebound of the tray and starter shaft back into the RPV, a latch is engaged. Finally, a head

  2. Deformation Characteristics and Sealing Performance of Metallic O-rings for a Reactor Pressure Vessel

    OpenAIRE

    Shen, Mingxue; PENG, Xudong; Xie, Linjun; Meng, Xiangkai; Li, Xinggen

    2016-01-01

    This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV). A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the ...

  3. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  4. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  5. Flow analyses for the LAVA-ERVC experiment and the KSNP under the external reactor vessel cooling using RELAP5/MOD3 code

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung-Ho; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik

    2005-01-01

    Flow analyses were performed using RELAP5/MOD3 code to investigate and verify the steam binding phenomena in the LAVA-ERVC experiment and to investigate the occurrence and the effects of steam binding for the KSNP under the external reactor vessel cooling. Flow analyses for the LAVA-ERVC experiments confirmed the steam binding occurrence in case of the limited steam venting and represented the LAVA-ERVC experimental results quite well. The flow analyses results for the KSNP under the external reactor vessel cooling address that water ingression and steam ventilation through the insulator are crucial factors determining the effective cool down via boiling heat removal at the outer surface of the RPV lower plenum. The flow analyses results for the base cases of the SBO and the 9.6 inch LBLOCA imply that the limited steam venting through the insulator induced the steam binding and eventually prevented the effective cooling at the outer surface of the RPV lower plenum. From the sensitivity study on the additional flow area for the steam venting, it could be found that the RPV lower plenum experienced effective cooling by smooth water circulation. The current RELAP5 flow analyses results for the KSNP under the external reactor vessel cooling address that prevention of steam binding phenomena should be settled first for the in-vessel corium retention through the external reactor vessel cooling. Implementation of additional flow path for the effective steam ventilation is highly recommended as one of the most promising countermeasures to enhance the coolability through the external reactor vessel cooling.

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

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Williams, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, B. Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Klasky, Hilda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

  7. 46 CFR 154.188 - Membrane tank: Inner hull steel.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Membrane tank: Inner hull steel. 154.188 Section 154.188... Structure § 154.188 Membrane tank: Inner hull steel. For a vessel with membrane tanks, the inner hull... “Rules for Building and Classing Steel Vessels”, 1981. ...

  8. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR

    Directory of Open Access Journals (Sweden)

    Flaspoehler Timothy

    2016-01-01

    Full Text Available One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV. While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR Power Plant, including DPA (displacements per atom and fast neutron fluence (>1 MeV and >0.1MeV. I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling methodology to bias a fixed-source MC (Monte Carlo simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  9. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Science.gov (United States)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  10. Keeping control when cutting through a reactor vessel

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1997-09-01

    UK Robotics` Advanced Teleoperation Controller (ATC) is a key component of one of the most extensive remote handling operations currently being undertaken - the removal of 165 mm diameter, 90 mm thick samples of carbon-manganese steel from the base of the Trawsfyndd reactor pressure vessel. These will then be used to assess the material properties of the vessel welds. (author).

  11. Steel making

    CERN Document Server

    Chakrabarti, A K

    2014-01-01

    "Steel Making" is designed to give students a strong grounding in the theory and state-of-the-art practice of production of steels. This book is primarily focused to meet the needs of undergraduate metallurgical students and candidates for associate membership examinations of professional bodies (AMIIM, AMIE). Besides, for all engineering professionals working in steel plants who need to understand the basic principles of steel making, the text provides a sound introduction to the subject.Beginning with a brief introduction to the historical perspective and current status of steel making together with the reasons for obsolescence of Bessemer converter and open hearth processes, the book moves on to: elaborate the physiochemical principles involved in steel making; explain the operational principles and practices of the modern processes of primary steel making (LD converter, Q-BOP process, and electric furnace process); provide a summary of the developments in secondary refining of steels; discuss principles a...

  12. Effect of Multi-Layered Corium Formations on Integrity of Steel Components under Steam Explosion Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hyun; Kim, Tae Hyun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of); Cho, Yong-Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The object of the present study is to examine effect of multi-layered corium formations on the integrity of steel components under a representative steam explosion condition. In this context, multi-layered corium formation conditions are assumed based on a previous study. Subsequently, stress evaluation of steel components is performed by TNT (trinitrotoluene) model for the steam explosion analysis and their results are discussed. In this paper, comparative numerical analyses were carried out to examine effect of the multi-layered corium formations on integrity of steel components under a typical steam explosion condition and the following conclusions were derived. (1) The highest maximum von Mises stress was calculated at RPV. However, stress values of all components did not exceed their yield strengths. (2) Effect of the 3-layer corium formation was higher than 2-layer corium formation. Resulting von Mises stress increased 20% than that of no corium formation and 16% than that of 2-layer corium formation.

  13. Vessel Operator System

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

  14. Design of a supercritical water-cooled reactor. Pressure vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Kai

    2008-08-15

    The High Performance Light Water Reactor (HPLWR) is a light water reactor with supercritical steam conditions which has been investigated within the 5th Framework Program of the European Commission. Due to the supercritical pressure of 25 MPa, water, used as moderator and as coolant, flows as a single phase through the core and can be directly fed to the turbine. Using the technology of coal fired power plants with supercritical steam conditions, the heat-up in the core is done in several steps to achieve the targeted high steam outlet temperature of 500.C without exceeding available cladding material limits. Based on a first design of a fuel assembly cluster for a HPLWR with a single pass core, the surrounding internals and the reactor pressure vessel (RPV) are dimensioned for the first time, following the safety standards of the nuclear safety standards commission in Germany. Furthermore, this design is extended to the incorporation of core arrangements with two and three passes. The design of the internals and the RPV are verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Additionally, a passive safety component for the feedwater inlet of the RPV of the HPLWR is designed. Its purpose is the reduction of the mass flow rate in case of a LOCA for a feedwater line break until further steps are executed. Starting with a simple vortex diode, several steps are executed to enhance the performance of the diode and adapt it to this application. Then, this first design is further optimized using combined 1D and 3D flow analyses. Parametric studies determine the performance and characteristic for changing mass flow rates for this backflow limiter. (orig.)

  15. Visual inspection of vessel internals; Visuelle Inspektion von Kerneinbauten

    Energy Technology Data Exchange (ETDEWEB)

    Rabe, G. [Siemens AG KWU, Erlangen (Germany)

    1999-08-01

    Visual inspection has matured to a qualified testing method and has become a standard method for inspection of reactor pressure vessels. Until today, all known defects in RPV internals have been detected by visual inspection. The codes KTA 3204 and DIN 25435-4 describe the framework conditions and requirements for visual inspections, which should be adhered to to the most possible extent. Visual inspections are carried by now at all RPV internals, also at those where access is difficult and limited. The inspection robot SUSI is applied in most cases. The camera and manipulator technology meanwhile has been upgraded to a standard performance quality allowing reliable, fast and easy visual inspection. The personnel is trained accordingly, so as to keep abreast with enhancements. Qualification of the inspection system has been simplified and standardised to a large extent. (orig/CB) [Deutsch] Die Sichtpruefung ist zu einem qualifizierten Pruefverfahren gereift und hat bei der Inspektion der RDB-Einbauten einen festen Platz eingenommen. Bisher wurden alle bekannten Schaeden an den RDB-Einbauten bei der Sichtpruefung festgestellt. In der KTA 3204 und der DIN 25435-4 sind die Rahmenbedingungen und Anforderungen an die Sichtpruefung beschrieben, die es gilt, weitestgehend einzuhalten. Mittlerweile werden an allen RDB-Einbauten, auch an den nur bedingt zugaenglichen, Sichtpruefungen vorgenommen. Dabei hat das Inspektionsfahrzeug SUSI inzwischen den breitesten Raum eingenommen. Die Entwicklung der Kamera- und Manipulatortechnik hat inzwischen einen Stand erreicht, der eine sichere, schnelle und einfache Sichtpruefung zulaesst. Das Pruefpersonal wird laufend fuer die Sichtpruefung geschult und qualifiziert. Die Qualifizierung des Inspektionssystems wurde weitestgehend vereinfacht und standardisiert. (orig.)

  16. Stress analysis of the reactor pressure vessel of the high performance light water reactors (HPLWR); Festigkeitsanalyse fuer den Reaktordruckbehaelter des High Performance Light Water Reactor (HPLWR)

    Energy Technology Data Exchange (ETDEWEB)

    Guelton, E.; Fischer, K.

    2006-12-15

    The High Performance Light Water Reactor (HPLWR) is one of the concepts of the Generation IV program. The main difference compared to current Light Water Reactors (LWR) results from the supercritical steam condition of the coolant. Due to the supercritical pressure of 25 MPa, water, used as moderator and coolant, flows as a single phase through the core. The temperatures at the outlet are above 500 C. These conditions have a major impact on the design of the Reactor Pressure Vessel (RPV). For the modelling a RPV concept is proposed, which resembles the design of current LWR and allows the use of approved materials on one side and also meets the additional demands on the other side. A first dimensioning of the RPV wall thicknesses and the geometrical proportions has been performed using the german KTA-guidelines. To verify these results, a stress analysis using the finite element method has been performed with the program ANSYS. The combined mechanical and thermal calculations provide the primary, secondary and peak stresses which are evaluated using the KTA-guidelines design loading (Level 0) and service loading level A for the different components. The results confirm the wall thicknesses estimated by Fischer et al. (2006), but there are peak stresses in the vicinity of the inlet and outlet flanges, which are very close to the allowed design limit. For larger diameters of the RPV those regions will become critical and the stresses might exceed the design limits. Design optimizations for those regions are proposed and evaluated. A readjusted geometry of the inlet flange reduces those stresses by 65%. (orig.)

  17. Hygro-Thermo-Mechanical Analysis of a Reactor Vessel

    Directory of Open Access Journals (Sweden)

    Jaroslav Kruis

    2012-01-01

    Full Text Available Determining the durability of a reactor vessel requires a hygro-thermo-mechanical analysis of the vessel throughout its service life. Damage, prestress losses, distribution of heat and moisture and some other quantities are needed for a durability assessment. A coupled analysis was performed on a two-level model because of the huge demands on computer hardware. This paper deals with a hygro-thermo-mechanical analysis of a reactor vessel made of prestressed concrete with a steel inner liner. The reactor vessel is located in Temelín, Czech Republic.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

  19. Deformation characteristics and sealing performance of metallic-O-ring for a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ming Xue; Peng, Xudong; Xie, Linjun; Meng, Xiang Kai [Engineering Research Center of Process Equipment and Its Remanufacture, Ministry of Education, Zhejiang University of Technology, Hangzhou (China); Li, Xing Gen [Ningbo Tiansheng Sealing Packing Co., Ltd., Ningbo (China)

    2016-04-15

    This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV). A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap) have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

  20. Deformation Characteristics and Sealing Performance of Metallic O-rings for a Reactor Pressure Vessel

    Directory of Open Access Journals (Sweden)

    Mingxue Shen

    2016-04-01

    Full Text Available This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV. A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

  1. The impact of mobile point defect clusters in a kinetic model of pressure vessel embrittlement

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1998-05-01

    The results of recent molecular dynamics simulations of displacement cascades in iron indicate that small interstitial clusters may have a very low activation energy for migration, and that their migration is 1-dimensional, rather than 3-dimensional. The mobility of these clusters can have a significant impact on the predictions of radiation damage models, particularly at the relatively low temperatures typical of commercial, light water reactor pressure vessels (RPV) and other out-of-core components. A previously-developed kinetic model used to investigate RPV embrittlement has been modified to permit an evaluation of the mobile interstitial clusters. Sink strengths appropriate to both 1- and 3-dimensional motion of the clusters were evaluated. High cluster mobility leads to a reduction in the amount of predicted embrittlement due to interstitial clusters since they are lost to sinks rather than building up in the microstructure. The sensitivity of the predictions to displacement rate also increases. The magnitude of this effect is somewhat reduced if the migration is 1-dimensional since the corresponding sink strengths are lower than those for 3-dimensional diffusion. The cluster mobility can also affect the evolution of copper-rich precipitates in the model since the radiation-enhanced diffusion coefficient increases due to the lower interstitial cluster sink strength. The overall impact of the modifications to the model is discussed in terms of the major irradiation variables and material parameter uncertainties.

  2. Computational evaluation of the constraint loss on the fracture toughness of reactor pressure vessel steels; Evaluacion computacional del efecto de la perdida de constriccion en la tenacidad de fractura de la vasija de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Serrano Garcia, M.

    2007-07-01

    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{sub 0} is obtained by testing pre-cracked Charpy specimens. The reason is that the T{sub 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{sub 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{sub 0} value obtained by testing the Charpy specimens included in the surveillance capsules. (Author) 111 ref.

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

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

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

  4. BIOASSAY VESSEL FAILURE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Vormelker, P

    2008-09-22

    Two high-pressure bioassay vessels failed at the Savannah River Site during a microwave heating process for biosample testing. Improper installation of the thermal shield in the first failure caused the vessel to burst during microwave heating. The second vessel failure is attributed to overpressurization during a test run. Vessel failure appeared to initiate in the mold parting line, the thinnest cross-section of the octagonal vessel. No material flaws were found in the vessel that would impair its structural performance. Content weight should be minimized to reduce operating temperature and pressure. Outer vessel life is dependent on actual temperature exposure. Since thermal aging of the vessels can be detrimental to their performance, it was recommended that the vessels be used for a limited number of cycles to be determined by additional testing.

  5. BY FRUSTUM CONFINING VESSEL

    Directory of Open Access Journals (Sweden)

    Javad Khazaei

    2016-09-01

    Full Text Available Helical piles are environmentally friendly and economical deep foundations that, due to environmental considerations, are excellent additions to a variety of deep foundation alternatives available to the practitioner. Helical piles performance depends on soil properties, the pile geometry and soil-pile interaction. Helical piles can be a proper alternative in sensitive environmental sites if their bearing capacity is sufficient to support applied loads. The failure capacity of helical piles in this study was measured via an experimental research program that was carried out by Frustum Confining Vessel (FCV. FCV is a frustum chamber by approximately linear increase in vertical and lateral stresses along depth from top to bottom. Due to special geometry and applied bottom pressure, this apparatus is a proper choice to test small model piles which can simulate field stress conditions. Small scale helical piles are made with either single helix or more helixes and installed in fine grained sand with three various densities. Axial loading tests including compression and tension tests were performed to achieve pile ultimate capacity. The results indicate the helical piles behavior depends essentially on pile geometric characteristics, i.e. helix configuration and soil properties. According to the achievements, axial uplift capacity of helical model piles is about equal to usual steel model piles that have the helixes diameter. Helical pile compression bearing capacity is too sufficient to act as a medium pile, thus it can be substituted other piles in special geoenvironmental conditions. The bearing capacity also depends on spacing ratio, S/D, and helixes diameter.

  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. Business Intelligence for Strategic Steel Constructions Sourcing

    DEFF Research Database (Denmark)

    Adeyemi, Oluseyi

    2010-01-01

    markets, government support for industry and stability of government}, to source steel constructions strategically. I undertook this project as a consultation for JB Contractors A/S {JBC} now referred to as Strongstaal A/S. JBC builds on its core competences in steel constructions, forgings, pressure...... vessels, welding, machining, heat treatment, corrosive treatment and quality control. It uses these core competencies to manufacture heavy duty, labour-intensive welded and machine processed steel structures in Eastern Europe. It has many years of sound project management experience and has enjoyed great...

  8. Tool steels

    DEFF Research Database (Denmark)

    Højerslev, C.

    2001-01-01

    On designing a tool steel, its composition and heat treatment parameters are chosen to provide a hardened and tempered martensitic matrix in which carbides are evenly distributed. In this condition the matrix has an optimum combination of hardness andtoughness, the primary carbides provide...... resistance against abrasive wear and secondary carbides (if any) increase the resistance against plastic deformation. Tool steels are alloyed with carbide forming elements (Typically: vanadium, tungsten, molybdenumand chromium) furthermore some steel types contains cobalt. Addition of alloying elements...... serves primarily two purpose (i) to improve the hardenabillity and (ii) to provide harder and thermally more stable carbides than cementite. Assuming proper heattreatment, the properties of a tool steel depends on the which alloying elements are added and their respective concentrations....

  9. Guam Abandoned Vessel Inventory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Guam. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  10. Florida Abandoned Vessel Inventory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Florida. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  11. Vessel Arrival Info - Legacy

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Vessel Arrival Info is a spreadsheet that gets filled out during the initial stage of the debriefing process by the debriefer. It contains vessel name, trip...

  12. 46 CFR 54.25-25 - Welding of quenched and tempered steels (modifies UHT-82).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Welding of quenched and tempered steels (modifies UHT-82... ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-25 Welding of quenched and tempered steels (modifies UHT-82). (a) The qualification of welding procedures, welders, and...

  13. ALICE HMPID Radiator Vessel

    CERN Multimedia

    2003-01-01

    View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

  14. Development of Integrated Regulatory Aging Management System related to Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Hanok; Park, Jeongsoon; Kim, Seonjae; Jhung, Myungjo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The primary function of the reactor vessel internals (RVIs) is to support the core, the control rod assemblies, the core support structure and the reactor pressure vessel (RPV) surveillance capsules. The RVIs have the additional function to direct the flow of the reactor coolant and provide shielding for the RPV. Ageing mechanisms are specific processes that gradually change characteristics of a component with time and use. According to the Generic Aging Lessons Learned (GALL) report, aging mechanisms, such as fatigue, embrittlement, corrosion, wear, radiation induced creep, relaxation and swelling, is related to RVIs. Establishing that effects of aging degradation in RVIs are adequately managed is vital for assuring continued functionality of RVIs. To achieve this goal, it is necessary to develop the regulatory standard as well as generic inspection and evaluation guideline for RVIs. In this paper, the Integrated Regulatory Aging Management System (IR-Aging), which efficiently manages key data necessary to the development of regulatory standards and assists effective evaluation of RVIs, is proposed. By using the proposed system, experts in different fields can co-operate to resolve safety issues and all users can share information and create valuable knowledge-base. In this paper, the Integrated Regulatory Aging Management System (IR-Aging) is proposed in order to manage data necessary to the development of regulatory standards and assists effective evaluation of RVIs. The proposed system provides various documents, such as US NRC and domestic regulatory documents, licensee's documents submitted to a regulatory body, and research documents. By using the proposed system, experts in different fields can co-operate to resolve safety issues and all users can share information and create valuable knowledge-base.

  15. Modular scaffolding for assembling the inside of an LNG vessel

    Energy Technology Data Exchange (ETDEWEB)

    Lienhard, R.W.

    1977-11-15

    A new scaffolding arrangement developed by Swiss Fabricating Inc., Pittsburgh, for finishing the inside of LNG vessels offers greater mobility and outrigger adjustability than conventional scaffolding and need not be specially constructed for each job. The scaffolding provides relatively large and open horizontal work areas without cross-bracing or tie rods. The structural steel base is supported from the bottom of the vessel by adjustable screw-jack supports. Adjustable outriggers can be extended to come close to the vessel's sides.

  16. Confinement Vessel Assay System: Design and Implementation Report

    Energy Technology Data Exchange (ETDEWEB)

    Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Mayo, Douglas R. [Los Alamos National Laboratory; Gomez, Cipriano D. [Retired CMR-OPS: OPERATIONS; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

    2012-07-18

    Los Alamos National Laboratory has a number of spherical confinement vessels remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1- to 2-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC&A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using {sup 252}Cf placed a various locations throughout the measurement system. Measurements were also performed with a {sup 252}Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

  17. A micro-mechanical analysis and an experimental characterisation of the behavior and the damaging processes of a 16MND5 pressure vessel steel at low temperature; Etude micromecanique et caracterisation experimentale du comportement et de l'endommagement de l'acier de cuve 16MND5 a basses temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Pesci, R

    2004-06-15

    As part of an important experimental and numerical research program launched by Electricite De France on the 16MND5 pressure vessel steel, sequenced and in-situ tensile tests are realized at low temperatures [-196 C;-60 C]. They enable to associate the observation of specimens, the complete cartography of which has been made with a scanning electron microscope (damaging processes, initiation and propagation of microcracks), with the stress states determined by X-ray diffraction, in order to establish relevant criteria. All these measurements enable to supply a two-scale polycrystalline modeling of behavior and damage (Mori-Tanaka/self-consistent) which is developed concurrently with the experimental characterization. This model proves to be a very efficient one, since it correctly reproduces the influence of temperature experimentally defined: the stress state in ferrite remains less important than in bainite (the difference never exceeds 150 MPa), whereas it is much higher in cementite. The heterogeneity of strains and stresses for each crystallographic orientation is well rendered; so is cleavage fracture normal to the {l_brace}100{r_brace} planes in ferrite (planes identified by electron back scattered diffraction during an in-situ tensile test at -150 C), which occurs sooner when temperature decreases, for a constant stress of about 700 MPa in this phase. (author)

  18. Design, Fabrication and Test Report on HANARO Instrumented Capsule (08M-01K) for the Evaluation of Irradiation Degradation of RPV Model Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Cho, M. S.; Kim, B. G.; Kang, Y. H.; Son, J. M.; Shin, Y. T.; Park, S. J.; Oh, S. Y.; Lee, J. H

    2009-03-15

    An instrumented capsule of 08M-01K was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of RPV model alloys. The basic structure of the 08M-01K capsule was based on the 07M-13N capsule which was successfully irradiated in the CT test hole of HANARO. 124 specimens such as PCVN, 1/2 PCVN, Charpy, small tensile, PA and TEM specimens of SA508 RPV model alloys were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe-Ag-Nb-Co neutron fluence monitors installed in the capsule. A friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated in the CT test hole of HANARO of a 30MW thermal output at 255{approx} 317 .deg. C up to a fast neutron fluence of 8.3x10{sup 19} n/cm{sup 2} (E>1.0 MeV). The obtained results will be very valuable for the evaluation on the extended operation of the KORI 1 Nuclear Power Reactor.

  19. Design, Fabrication and Test Report on Instrumented Capsule (08M-02K) for Irradiation Test of RPV Model Alloys in HANARO OR5 Test Hole

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Cho, M. S.; Kim, B. G.; Kang, Y. H.; Son, J. M.; Shin, Y. T.; Park, S. J.; Oh, S. Y

    2009-09-15

    An instrumented capsule of 08M-02K was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of RPV model alloys. The basic structure of the 08M-02K capsule was based on the 07M-21K capsule which was successfully irradiated in the OR5 test hole of HANARO. 228 specimens such as PCVN, 1/2 PCVN, Charpy, small tensile and TEM specimens of SA508 RPV model alloys were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. A friction welded tube between STS304 and AI1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during capsule cutting process in HANARO. The capsule was irradiated in the OR5 test hole of HANARO of a 30MW thermal output at 269{approx}310 .deg. C up to a fast neutron fluence of 4.4x10{sup 19}(n/cm{sup 2}) (E>1.0 MeV). The obtained results will be very valuable for the evaluation on the 2nd extended operation of the KORI 1 Nuclear Power Reactor.

  20. Design and analysis of multicavity prestressed concrete reactor vessels. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

  1. Development of structural steels for nuclear application

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Chi, S. H.; Ryu, W. S.; Lee, B. S.; Kim, D. H.; Kim, J. H.; Oh, Y. J.; Byun, T. S.; Yoon, J. H.; Park, D. K.; Oh, J. M.; Cho, H. D.; Kim, H.; Kim, H. D.; Kang, S. S.; Kim, J. W.; Ahn, S. B.

    1997-08-01

    To established the bases of nuclear structural material technologies, this study was focused on the localization and improvement of nuclear structural steels, the production of material property data, and technology developments for integrity evaluation. The important test and analysis technologies for material integrity assessment were developed, and the materials properties of the pressure vessel steels were evaluated systematically on the basis of those technologies, they are microstructural characteristics, tensile and indentation deformation properties, impact properties, and static and dynamic fracture toughness, fatigue and corrosion fatigue etc. Irradiation tests in the research reactors were prepared or completed to obtain the mechanical properties of irradiated materials. The improvement of low alloy steel was also attempted through the comparative study on the manufacturing processes, computer assisted alloy and process design, and application of the inter critical heat treatment. On the other hand, type 304 stainless steels for reactor internals were developed and tested successfully. High strength type 316LN stainless steels for reactor internals were developed and the microstructural characteristics, corrosion resistance, mechanical properties at high temperatures, low cycle fatigue property etc. were tested and analyzed in the view point of the effect of nitrogen. Type 347 stainless steels with high corrosion resistance and toughness for pipings and tubes and low-activated Cr-Mn steels were also developed and their basic properties were evaluated. Finally, the martensitic stainless steels for turbine blade were developed and tests. (author). 242 refs., 100 tabs., 304 figs.

  2. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  3. 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 ...

  4. ITER Vacuum Vessel design and construction

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Jones, L. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Jun, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC ' Sintez' , Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector -25, Gandhinagar 382025 (India); Preble, J.; Reich, J. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2012-08-15

    After implementing a few design modifications (referred to as the 'Modified Reference Design') in 2009, the Vacuum Vessel (VV) design had been stabilized. The VV design is being finalized, including interface components such as support rails and feedthroughs for the in-vessel coils. It is necessary to make adjustments to the locations of the blanket supports and manifolds to accommodate design modifications to the in-vessel coils. The VV support design is also being finalized considering a structural simplification. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. The detailed layout of ferritic steel plates and borated steel plates was optimized based on the toroidal field ripple analysis. A dynamic test on the inter-modular key to support the blanket modules was performed to measure the dynamic amplification factor (DAF). An R and D program has started to select and qualify the welding and cutting processes for the port flange lip seal. The ITER VV material 316 L(N) IG was already qualified and the Modified Reference Design was approved by the Agreed Notified Body (ANB) in accordance with the Nuclear Pressure Equipment Order procedure.

  5. Inspection findings in austenitic RPV internals of German BWR plants and BWRs built in other countries and resulting measures for Isar 1 nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Erve, M.; Hurlebaus, D.; Marschke, D.; Senski, G. [Siemens AG, Erlangen (Germany). Power Generation Group; Maier, V. [Bayernwerk Kernenergie GmbH, Muenchen (Germany); Baeumler, H.-J.; Winter, F. [TUeV Anlagen-und Umwetttechnik GmbH, Muenchen (Germany)

    1999-06-01

    As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Wuergassen Nuclear Power Station need not be feared. Although the material and corrosion-chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert`s agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept. (orig.) 18 refs.

  6. German RPV safety assessment. Underpinning of the procedure by complementary test results measured in the hot cells; Der deutsche RDB-Sicherheitsnachweis. Untermauerung der Vorgehensweise durch ergaenzende Kennwertermittlung in den Heissen Zellen

    Energy Technology Data Exchange (ETDEWEB)

    Keim, Elisabeth; Hein, Hieronymus; Gundermann, Arnulf [AREVA NP GmbH (Germany); Hoffmann, Harald [VGB (Germany); Koenig, Guenter; Ilg, Ulf [EnBW (Germany); Nagel, Gerhard [e-on Kernkraft (Germany); Widera, Martin [RWE (Germany); Rebsamen, Daniel [KKW Goesgen (Germany)

    2008-07-01

    In Germany the assessment of the RPV (reactor pressure vessel) integrity is regulated by the German Code KTA 3201.2, based on a deterministic concept. The material characteristics are indexed by the reference temperature RTNDT which is determined by mechanical tests. The comparison with fracture mechanical characteristics shows that the fracture toughness curve KIc (T-RTNDT) is an envelope of the experimental data. Worldwide a tendency is observed to implement besides the established RTNDT concept another concept based on fracture mechanical characteristics. The advantage of the new concept is a direct determination of the ductile-brittle transition temperature using fracture mechanical tests, which is supposed to allow a more realistic transferability to the component. In order to integrate the Master-curve-concept into the German standards several questions have to be answered: for instance the relation to a representative data base of irradiated German RPV materials and the influence of the specimen shape and size. There is still a necessity to compare the established concept with the new concept und to clarify whether crack arrest curves of irradiated materials could be assessed and integrated into the master curve concept. Within the project CARISMA a data base of fracture toughness values of irradiated original RPV materials representative for all four German PWR generations was compiled. The RTNDT and the master-curve concept were used for the evaluation of the generic data in order to allow the comparison of both concepts. The main results are the following: The lower-bound ASME KIc curve for crack initiation (brittle failure) was confirmed by the measured fracture toughness data of the irradiated materials. The significant influence of copper and nickel on the irradiation behaviour of RPC materials was confirmed. The transition temperature shifts ?T41 and ?T0 show relatively good correlation. Fracture mechanical specimens type SE(B) 10 mm x 10 mm are

  7. Hegelian Steel

    DEFF Research Database (Denmark)

    Kjær, Poul F.

    2015-01-01

    . Developing a Hegelian inspired historical-sociological approach this paper however argues that national and transnational societies emerged simultaneously and in a co-evolutionary and mutually supportive fashion. In most European settings national societies did not become the central horizon of individuals...... of the European steel industry....

  8. Maury Journals - German Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — German vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  9. Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    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/sup 0/C (196/sup 0/F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa (18,700 psi).

  10. PRESSURE-RESISTANT VESSEL

    NARCIS (Netherlands)

    Beukers, A.; De Jong, T.

    1997-01-01

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

  11. Containment vessel drain system

    Energy Technology Data Exchange (ETDEWEB)

    Harris, Scott G.

    2018-01-30

    A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

  12. 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.

  13. 46 CFR 54.25-15 - Low temperature operation-high alloy steels (modifies UHA-23(b) and UHA-51).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Low temperature operation-high alloy steels (modifies... (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-15 Low temperature operation—high alloy steels (modifies UHA-23(b) and UHA-51). (a) Toughness...

  14. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  15. Switch to Stribild versus continuation of NVP or RPV with FTC and TDF in virologically suppressed HIV adults: a STRATEGY-NNRTI subgroup analysis

    Directory of Open Access Journals (Sweden)

    Hans-Juergen Stellbrink

    2014-11-01

    Full Text Available Introduction: Switch to Stribild (STB was non-inferior to continuation of a non-nucleoside reverse transcriptase inhibitor (NNRTI with emtricitabine and tenofovir DF (FTC/TDF at week 48 in virologically suppressed HIV adults (1. We report the Week 48 efficacy and safety of STB versus nevirapine (NVP or rilpivirine (RPV with FTC/TDF in suppressed subjects. Materials and Methods: Virologically suppressed subjects on an NNRTI with FTC/TDF regimens for ≥6 months were randomized (2:1 to switch to STB versus continue their NNRTI regimen. Eligibility criteria included no documented resistance to FTC and TDF, no history of virologic failure and eGFR ≥70 mL/min. The primary endpoint was the proportion of subjects in the modified ITT population who maintained HIV-1 RNA <50 copies(c/mL at Week 48 by FDA snapshot algorithm (12% non-inferiority margin. Subgroup analysis by non-EFV NNRTI use (NVP [74]; RPV [19]; etravirine [3] at screening was pre-specified. Results: The mITT population included 433 subjects who were randomized and treated. In the non-EFV NNRTI subgroup, 59 switched to STB; 37 continued a non-EFV NNRTI (27 NVP, 10 RPV with FTC/TDF. At week 48, 97% STB versus 95% non-EFV NNRTI maintained HIV-1 RNA <50 c/mL. No emergent resistance was detected in either group. No difference in median increases from baseline in CD4 count at week 48 (cells/µL: 25 STB versus 55 non-EFV NNRTI (p=0.78. No discontinuation due to adverse events; no cases of proximal renal tubulopathy. As expected, there were no significant changes in the frequency of neuropsychiatric symptoms (i.e. anxiety, insomnia, dizziness, vivid dreams, weird/intense dreams, and nightmares reported on the HIV Symptom Index at week 48 compared to baseline after switching to STB. There was a greater but non-progressive decrease from baseline in eGFR in the STB versus non-EFV NNRTI group; median changes (mL/min at week 48: −9.1 versus −1.4. Switch to STB was associated with a higher

  16. 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.

  17. 2013 Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  18. 2011 Passenger Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  19. 2011 Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  20. 2013 Passenger Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  1. 2013 Tanker Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  2. 2013 Cargo Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  3. High Performance Marine Vessels

    CERN Document Server

    Yun, Liang

    2012-01-01

    High Performance Marine Vessels (HPMVs) range from the Fast Ferries to the latest high speed Navy Craft, including competition power boats and hydroplanes, hydrofoils, hovercraft, catamarans and other multi-hull craft. High Performance Marine Vessels covers the main concepts of HPMVs and discusses historical background, design features, services that have been successful and not so successful, and some sample data of the range of HPMVs to date. Included is a comparison of all HPMVs craft and the differences between them and descriptions of performance (hydrodynamics and aerodynamics). Readers will find a comprehensive overview of the design, development and building of HPMVs. In summary, this book: Focuses on technology at the aero-marine interface Covers the full range of high performance marine vessel concepts Explains the historical development of various HPMVs Discusses ferries, racing and pleasure craft, as well as utility and military missions High Performance Marine Vessels is an ideal book for student...

  4. Cheboygan Vessel Base

    Data.gov (United States)

    Federal Laboratory Consortium — Cheboygan Vessel Base (CVB), located in Cheboygan, Michigan, is a field station of the USGS Great Lakes Science Center (GLSC). CVB was established by congressional...

  5. Maury Journals - US Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — U.S. vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  6. 2011 Cargo Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  7. 2011 Tanker Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  8. 2013 Fishing Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  9. Coastal Logbook Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch (landed catch) and effort for fishing trips made by vessels that have been issued a Federal permit for the Gulf of Mexico reef fish,...

  10. LANL Robotic Vessel Scanning

    Energy Technology Data Exchange (ETDEWEB)

    Webber, Nels W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-25

    Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

  11. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mkrtchyan, Lilit; Schau, Henry [TUEV SUED Energietechnik GmbH, Mannheim (Germany). Abt. Strukturverhalten; Wolf, Werner; Holzer, Wieland [TUEV SUED Industrie Service GmbH, Muenchen (Germany). Abt. Behaelter und Turbosatz; Wernicke, Robert; Trieglaff, Ralf [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany). Abt. Festigkeit und Konstruktion

    2011-08-15

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  12. Enhancing supply vessel safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    A supply-vessel bridge installation consists of a navigating bridge and a control position aft, from which operators control the ship when close to rigs or platforms, and operate winches and other loading equipment. The international Convention for Safety of I Ale at Sea (SOLAS) does not regulate the layout, so design varies to a large degree, often causing an imperfect working environment. As for other types of ships, more than half the offshore service vessel accidents at sea are caused by bridge system failures. A majority can be traced back to technical design, and operational errors. The research and development project NAUT-OSV is a response to the offshore industry's safety concerns. Analysis of 24 incidents involving contact or collision between supply vessels and offshore installations owned or operated by Norwegian companies indicated that failures in the bridge system were often the cause.

  13. Oxidation effects during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Almyashev, V.I.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Sulatsky, A.A.; Vitol, S.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V. [Ioffe Institute, St. Petersburg (Russian Federation); Bechta, S. [Royal Institute of Technology (KHT), Stockholm (Sweden); Barrachin, M.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [Joint Research Centre, Institut für Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI (France)

    2016-08-15

    Highlights: • Corium–steel interaction tests were re-examined particularly for transient processes. • Oxidation of corium melt was sensitive to oxidant supply and surface characteristics. • Consequences for vessel steel corrosion rates in severe accidents were discussed. - Abstract: In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.

  14. Plasma nitriding of steels

    CERN Document Server

    Aghajani, Hossein

    2017-01-01

    This book focuses on the effect of plasma nitriding on the properties of steels. Parameters of different grades of steels are considered, such as structural and constructional steels, stainless steels and tools steels. The reader will find within the text an introduction to nitriding treatment, the basis of plasma and its roll in nitriding. The authors also address the advantages and disadvantages of plasma nitriding in comparison with other nitriding methods. .

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Inspection of dissimilar metal welds in reactor pressure vessels in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P. [Tecnatom, S.A., San Sebastian de los Reyes, Madrid (Spain)

    2011-07-01

    MRP-139 recommendations for inspection of dissimilar metal (DM) welds in PWR vessels were launched in the last years in the USA. Basically, it increases the frequency of the examinations in these type of welds, with major emphasis in the hot loops, adding one intermediate inspection at the ten years interval in outlet nozzles. The spanish nuclear power plants (NPP's) have begun the implementation of this type of inspections on the vessel nozzles DM welds. As this type of inspections could have an impact in the critical path duration of the outage, it is necessary the use of a mechanical equipment able to examine the nozzles DM welds in a short vessel occupation time (VOT) with high quality, qualified techniques and minimum requirements of the refuelling platform. Tecnatom undertook the design and development of a new more advanced equipment, named TENIS-DM, for implementing the reactor pressure vessel (RPV) nozzles examination. This equipment was designed in order to accomplish the stringent requirements and the updated examination techniques; it was used for the inspection of the DM welds of Asco 1 NPP inlet and outlet nozzles in March 2011. Examination techniques and procedures were qualified through the GRUVAL validation program, based on ENIC methodology. Mechanical scanner was equipped with a large number of examination probes, and TV cameras -for visual inspection and also for monitoring the ultrasonic inspections. A remote operated submarine was also used to give support to the operational personnel during the manipulation of the equipment and its movements from one nozzle to the others. During two months before the inspection, tests of the complete inspection system were made on a nozzle mock-up installed in a 4 meters deep well at Tecnatom's facilities; this scenario was also used during the training sessions of the inspection crew. The defined technical and practical objectives were achieved: use of qualified techniques and minimal impact on the

  17. Fatigue corrosion of pressure vessel steel 16 MND 5 exposed to primary PWR coolant: effect of passivation-depassivation phenomena at the tip of a crack. Fatigue-corrosion de l'acier de cuve 16 MND 5 en milieu primaire de REP: role des phenomenes de depassivation-repassivation en pointe de fissure

    Energy Technology Data Exchange (ETDEWEB)

    Combrade, P.; Foucault, M. (Institut de Recherches de la Siderurgie Francaise (IRSID), 78 - Saint-Germain-en-Laye (France))

    1992-10-01

    Kinetics of crack propagation under the coating of some PWR vats. The corrosive atmosphere acts more strongly on steels with high sulfur content. The role of sulfur in the cracking process of steel 16 MND 5 in primary cooling is explained. (Author). 7 refs., 3 figs.

  18. GOLD PRESSURE VESSEL SEAL

    Science.gov (United States)

    Smith, A.E.

    1963-11-26

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

  19. Network of endocardial vessels.

    Science.gov (United States)

    Lee, Byung-Cheon; Kim, Hong Bae; Sung, Baeckkyoung; Kim, Ki Woo; Sohn, Jamin; Son, Boram; Chang, Byung-Joon; Soh, Kwang-Sup

    2011-01-01

    Although there have been reports on threadlike structures inside the heart, they have received little attention. We aimed to develop a method for observing such structures and to reveal their ultrastructures. An in situ staining method, which uses a series of procedures of 0.2-0.4% trypan blue spraying and washing, was applied to observe threadlike structures on the surfaces of endocardia. The threadlike structures were isolated and observed by using confocal laser scanning microscopy (CLSM) and transmission electron microscopy (TEM). Networks of endocardial vessels (20 μm in thickness) with expansions (40-100 μm in diameter) were visualized; they were movable on the endocardium of the bovine atrium and ventricle. CLSM showed that (1) rod-shaped nuclei were aligned along the longitudinal direction of the endocardial vessel and (2) there were many cells inside the expansion. TEM on the endocardial vessel revealed that (1) there existed multiple lumens (1-7 μm in diameter) and (2) the extracellular matrices mostly consisted of collagen fibers, which were aligned along the longitudinal direction of the endocardial vessel or were locally organized in reticular structures. We investigated the endocardial circulatory system in bovine cardiac chambers and its ultrastructures, such as nucleic distributions, microlumens, and collagenous extracellular matrices. Copyright © 2011 S. Karger AG, Basel.

  20. Pressurized Vessel Slurry Pumping

    Energy Technology Data Exchange (ETDEWEB)

    Pound, C.R.

    2001-09-17

    This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air.

  1. Ultrahigh carbon steels, Damascus steels, and superplasticity

    Energy Technology Data Exchange (ETDEWEB)

    Sherby, O.D. [Stanford Univ., CA (United States). Dept. of Materials Science and Engineering; Wadsworth, J. [Lawrence Livermore National Lab., CA (United States)

    1997-04-01

    The processing properties of ultrahigh carbon steels (UHCSs) have been studied at Stanford University over the past twenty years. These studies have shown that such steels (1 to 2.1% C) can be made superplastic at elevated temperature and can have remarkable mechanical properties at room temperature. It was the investigation of these UHCSs that eventually brought us to study the myths, magic, and metallurgy of ancient Damascus steels, which in fact, were also ultrahigh carbon steels. These steels were made in India as castings, known as wootz, possibly as far back as the time of Alexander the Great. The best swords are believed to have been forged in Persia from Indian wootz. This paper centers on recent work on superplastic UHCSs and on their relation to Damascus steels. 32 refs., 6 figs.

  2. Tensile and Fatigue Behavior of ASS304 for Cold Stretching Pressure Vessels at Cryogenic Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon Seok [The 5th R and D Institute, Agency for Defense Development, Daejeon (Korea, Republic of); Kim, Jae Hoon; Na, Seong Hyun [Chungnam National Univ., Daejon (Korea, Republic of); Lee, Youn Hyung [Korean Gas Safety Corporation, Chungju (Korea, Republic of); Kim, Sung Hun [Daechang Solution Co. Ltd, Busan (Korea, Republic of); Kim, Young Kyun; Kim, Ki Dong [Korean Gas Corporation, R and D Division, Ansan (Korea, Republic of)

    2016-05-15

    Cold stretching(CS) pressure vessels from ASS304 (austenitic stainless steel 304) are used for the transportation and storage of liquefied natural gas(LNG). CS pressure vessels are manufactured by pressurizing the finished vessels to a specific pressure to produce the required stress σk. After CS, there is some degree of plastic deformation. Therefore, CS vessels have a higher strength and lighter weight compared to conventional vessels. In this study, we investigate the tensile and fatigue behavior of ASS304 sampled by CS pressure vessels in accordance with the ASME code at cryogenic temperature. From the fatigue test results, we show S-N curves using a statistical method recommended by JSEM-S002. We carried out the fractography of fractured specimens using scanning electron microscopy (SEM)

  3. Gamma radiography of refractory-lined vessels and components

    Energy Technology Data Exchange (ETDEWEB)

    Lapinski, N. P.

    1978-08-01

    Materials used in coal-conversion systems are exposed to high pressure, high temperature, corrosive and erosive gases, and liquids containing particulate matter. These severe environments necessitate an assessment of the integrity of components to prevent premature failures. Gamma radiography was evaluated as a viable technique for testing such components in the laboratory or after operation in situ. Penetrameters (image-quality indicators) were developed for refractory-lined vessels and transfer lines, and exposure times for various combinations of refractory-steel thicknesses were determined. Radiography with /sup 60/Co was performed on gasifier vessels, combustor vessels, and critical transfer lines in existing pilot plants using the experience gained through laboratory experiments. The results show that gamma radiography is a practical and effective method to detect critical conditions in coal-conversion system components. 18 figures, 3 tables.

  4. Hawaii Abandoned Vessel Inventory, Kauai

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Kauai. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  5. CNMI Abandoned Vessel Inventory, Tinian

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Tinian. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  6. Puerto Rico Abandoned Vessel Inventory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Puerto Rico. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  7. American Samoa Abandoned Vessel Inventory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for American Samoa. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  8. Hawaii Abandoned Vessel Inventory, Oahu

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Oahu, Hawaii. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  9. Hawaii Abandoned Vessel Inventory, Molokai

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Molokai, Hawaii. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  10. CNMI Abandoned Vessel Inventory, Rota

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Rota. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  11. Hawaii Abandoned Vessel Inventory, Lanai

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Lanai. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  12. For-Hire Vessel Directory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Vessel Directory is maintained as the sample frame for the For-Hire Survey. I contains data on for-hire vessels on the Atlantic and Gulf coasts. Data include...

  13. CNMI Abandoned Vessel Inventory, Saipan

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Saipan. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  14. Hawaii Abandoned Vessel Inventory, Maui

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Maui. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral habitats...

  15. Vessels in Transit - Web Tool

    Data.gov (United States)

    Department of Transportation — A web tool that provides real-time information on vessels transiting the Saint Lawrence Seaway. Visitors may sort by order of turn, vessel name, or last location in...

  16. Quest for steel quality: the role of metallurgical chemistry

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A. [Toronto Univ., ON (Canada). Dept. of Metallurgy and Materials Science

    2000-10-01

    Improvements in the quality of steels and the role played by metallurgical chemistry to bring about those improvements are discussed. The particular emphasis is on the chemical behaviour of solutes in molten steel and the reaction between steel, slag and refractory materials and the manner in which they influence the physical properties and performance of the steel product. As an illustration of the contribution of chemistry to steel making the case of the steel plates used in the construction of the Titanic is cited. In 1911 when the Titanic was constructed by Harland and Wolff at their Belfast shipyard, the steel plates used in the hull met all then current specifications. In 1992 when a number of steel samples recovered from the Titanic were examined, it was found that the hull of the vessel was constructed of low carbon, semi-killed steel, produced in the open-hearth process. Microstructural analysis showed extensive carbon banding, typical of hot rolled 0.2 per cent carbon steel. Also found were long manganese sulphide inclusions elongated in the rolling direction, some of which exceeded 25 mm in length. It was determined that as a consequence of these inclusions, at a seawater temperature of 0 degree C, the hull plates of the Titanic had essentially no resistance to fracture. Today's high quality steels used in applications such as Arctic pipelines, offshore platforms, icebreakers and ships for the transportation of natural gas, oxygen and sulphur concentrations are frequently less than 10 ppm. These elements have a profound influence of the quality of the final steel products by virtue of their effect of hindering the formation of inclusions. 2 refs., 3 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-18

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

  18. Pressure vessel design manual

    Energy Technology Data Exchange (ETDEWEB)

    Moss, D.R.

    1987-01-01

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

  19. New research vessels

    Science.gov (United States)

    1984-04-01

    Two “new” ocean-going research vessels operated by the Scripps Institution of Oceanography and the National Science Foundation (NSF) will soon begin full-time scientific duties off the coast of California and in the Antarctic, respectively. The 37.5-m Scripps vessel, named Robert Gordon Sprout in honor of the ex-president of the University of California, replaces the smaller ship Ellen B. Scripps, which had served the institution since 1965. The new ship is a slightly modified Gulf Coast workboat. Under the name of Midnight Alaskan, it had been used for high-resolution geophysical surveys in American and Latin American waters by such firms as Arco Oil & Gas, Exxon, Pennzoil, and Racal-Decca before its purchase by Scripps from a Lousiana chartering firm last summer.

  20. Large vessel vasculitides

    OpenAIRE

    Morović-Vergles, Jadranka; Pukšić, Silva; Gudelj Gračanin, Ana

    2013-01-01

    Large vessel vasculitis includes Giant cell arteritis and Takayasu arteritis. Giant cell arteritis is the most common form of vasculitis affect patients aged 50 years or over. The diagnosis should be considered in older patients who present with new onset of headache, visual disturbance, polymyalgia rheumatica and/or fever unknown cause. Glucocorticoides remain the cornerstone of therapy. Takayasu arteritis is a chronic panarteritis of the aorta ant its major branches presenting commonly in y...

  1. Very Versatile Vessel

    Science.gov (United States)

    2009-09-01

    data. This source provides information on aluminum hydrofoil vessels without the added weight of foil structures. The composite armor around the...seating compartment. The sides should also limit wave splash on the deck. The freeboard should contribute reserve buoyancy , increasing large-angle and...Resistance, Powering, and Propulsion Savitsky’s Method Since model testing data or other reliable performance data was unavailable for the proposed

  2. Evaluation of Failure Probability of BWR Vessel Under Cool-down and LTOP Transient Conditions Using PROFAS-RV PFM Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong-Min; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The round robin project was proposed by the PFM Research Subcommittee of the Japan Welding Engineering Society to Asian Society for Integrity of Nuclear Components (ASINCO) members, which is designated in Korea as Phase 2 of A-Pro2. The objective of this phase 2 of RR analysis is to compare the scheme and results related to the assessment of structural integrity of RPV for the events important to safety in the design consideration but relatively low fracture probability. In this study, probabilistic fracture mechanics analysis was performed for the round robin cases using PROFAS-RV code. The effects of key parameters such as different transient, fluence level, Cu and Ni content, initial RT{sub NDT} and RT{sub NDT} shift model on the failure probability were systematically compared and reviewed. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel.

  3. Vessel Traffic Services.

    Science.gov (United States)

    1982-12-01

    Yorker" articles titled Silent Spring by Rachel Carson in 1963 produced a unifying effect, "the sort of rallying point of the movement to protect the...6232, 92d Cong., 1st. sess., 1971, p. 2. 15. Carson , Rachel L. , The Sea Around Us, New York: Oxford Univesity Press, 195-, p. IV. 16. U.S., Congress...Government Printing Office, 1974. 63. Buhler, L. and Geiger, J., Vessel Traffic Data Extraction MethodoloqX, Silver Spring , Maryland, O6erFae-tns

  4. Vanishing corneal vessels

    Science.gov (United States)

    Nicholson, Luke; Chana, Rupinder

    2013-01-01

    We wish to highlight the importance of acknowledging the accompanying effects of topical phenylephrine drops on the eye other than its intended mydriasis. We reported a case of a 92-year-old woman with a corneal graft who was noted to have superficial corneal vascularisation which was not documented previously. After the instillation of topical tropicamide 1% and phenylephrine 2.5%, for funduscopy, the corneal vascularisation was not visible. When reassessed on another visit, tropicamide had no effect on the vessels and only phenylephrine did. We wish to highlight that when reviewing patients in cornea clinics, instilling phenylephrine prior to being seen may mask important corneal vascularisation. PMID:24121816

  5. The steel scrap age.

    Science.gov (United States)

    Pauliuk, Stefan; Milford, Rachel L; Müller, Daniel B; Allwood, Julian M

    2013-04-02

    Steel production accounts for 25% of industrial carbon emissions. Long-term forecasts of steel demand and scrap supply are needed to develop strategies for how the steel industry could respond to industrialization and urbanization in the developing world while simultaneously reducing its environmental impact, and in particular, its carbon footprint. We developed a dynamic stock model to estimate future final demand for steel and the available scrap for 10 world regions. Based on evidence from developed countries, we assumed that per capita in-use stocks will saturate eventually. We determined the response of the entire steel cycle to stock saturation, in particular the future split between primary and secondary steel production. During the 21st century, steel demand may peak in the developed world, China, the Middle East, Latin America, and India. As China completes its industrialization, global primary steel production may peak between 2020 and 2030 and decline thereafter. We developed a capacity model to show how extensive trade of finished steel could prolong the lifetime of the Chinese steelmaking assets. Secondary steel production will more than double by 2050, and it may surpass primary production between 2050 and 2060: the late 21st century can become the steel scrap age.

  6. The issues of endurance when docking vessels in floating docks (review)

    OpenAIRE

    ANTONENKO SERGEY

    2017-01-01

    Large capacity docks are relatively infrequently loaded with commensurate vessels. In this context, one of the ways to enhance the operational efficiency of docks is handling ships in groups. The mechanical stress-strain state of the structure including a steel pontoon dock with vessels in it is hard to calculate and it may entail failures in its endurance. Of particular interest is the case of a large-capacity vessel and several small-displacement ones making up a dock group. Large displ...

  7. 46 CFR 289.2 - Vessels included.

    Science.gov (United States)

    2010-10-01

    ... CONSTRUCTION-DIFFERENTIAL SUBSIDY VESSELS, OPERATING-DIFFERENTIAL SUBSIDY VESSELS AND OF VESSELS SOLD OR ADJUSTED UNDER THE MERCHANT SHIP SALES ACT 1946 § 289.2 Vessels included. Vessels subject to the provisions of this part are: (a) All vessels which may in the future be constructed or sold with construction...

  8. Blood flow reprograms lymphatic vessels to blood vessels.

    Science.gov (United States)

    Chen, Chiu-Yu; Bertozzi, Cara; Zou, Zhiying; Yuan, Lijun; Lee, John S; Lu, MinMin; Stachelek, Stan J; Srinivasan, Sathish; Guo, Lili; Vicente, Andres; Vincente, Andres; Mericko, Patricia; Levy, Robert J; Makinen, Taija; Oliver, Guillermo; Kahn, Mark L

    2012-06-01

    Human vascular malformations cause disease as a result of changes in blood flow and vascular hemodynamic forces. Although the genetic mutations that underlie the formation of many human vascular malformations are known, the extent to which abnormal blood flow can subsequently influence the vascular genetic program and natural history is not. Loss of the SH2 domain-containing leukocyte protein of 76 kDa (SLP76) resulted in a vascular malformation that directed blood flow through mesenteric lymphatic vessels after birth in mice. Mesenteric vessels in the position of the congenital lymphatic in mature Slp76-null mice lacked lymphatic identity and expressed a marker of blood vessel identity. Genetic lineage tracing demonstrated that this change in vessel identity was the result of lymphatic endothelial cell reprogramming rather than replacement by blood endothelial cells. Exposure of lymphatic vessels to blood in the absence of significant flow did not alter vessel identity in vivo, but lymphatic endothelial cells exposed to similar levels of shear stress ex vivo rapidly lost expression of PROX1, a lymphatic fate-specifying transcription factor. These findings reveal that blood flow can convert lymphatic vessels to blood vessels, demonstrating that hemodynamic forces may reprogram endothelial and vessel identity in cardiovascular diseases associated with abnormal flow.

  9. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures and...

  10. Creep strength of N9 and N10 material (steel)

    Energy Technology Data Exchange (ETDEWEB)

    1943-02-17

    This letter was a response to the receipt of tables of information from the materials-testing laboratory at Ludwigshafen. The tables dealt with various properties of N9 and N10 steels for production of high-pressure hydrogenation vessels. The letter expressed questions about some of the information, especially about the methods of tempering the test steels and about certain figures for contraction of N9. The letter gave Leuna's values for creep strength (long-time rupture strength) after 20,000 hours of operation as 11 to 15 kg/mm/sup 2/ for N9 versus 26 to 30 kg/mm/sup 2/ for N10, and said that similar relationships existed in values for continuous creep strength (fatigue strength for an infinite time) between the steels. It had generally been Leuna's experience in high-temperature ruptures of pipes, though, that long before brittleness and contraction had set in very much, the physical action of hydrogen on the steel had led to ruptures. Because of this hydrogen activity, it was the aim of current work to increase the stability of N10 against hydrogen. One reason for the effort was to avoid being forced to rely on austenitic steels alone for the future development of the best steels for pressure vessels, since the resulting large demands on chromium and manganese might not always be able to be supplied. It was known that stability against hydrogen could be increased by addition of titanium to the steel, but it was not known to what extent creep strength in a hydrogen atmosphere could be improved thereby. Addition of titanium could also allow a corresponding reduction in the usage of vanadium or tungsten. Further discussions with steel suppliers were recommended.

  11. The vessel fluence; Fluence cuve

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This book presents the proceedings of the technical meeting on the reactors vessels fluence. They are grouped in eight sessions: the industrial context and the stakes of the vessels control; the organization and the methodology for the fluence computation; the concerned physical properties; the reference computation methods; the fluence monitoring in an industrial context; vessels monitoring under irradiation; others methods in the world; the research and development programs. (A.L.B.)

  12. Design of Semi-composite Pressure Vessel using Fuzzy and FEM

    Science.gov (United States)

    Sabour, Mohammad H.; Foghani, Mohammad F.

    2010-04-01

    The present study attempts to present a new method to design a semi-composite pressure vessel (known as hoop-wrapped composite cylinder) using fuzzy decision making and finite element method. A metal-composite vessel was designed based on ISO criteria and then the weight of the vessel was optimized for various fibers of carbon, glass and Kevlar in the cylindrical vessel. Failure criteria of von-Mises and Hoffman were respectively employed for the steel liner and the composite reinforcement to characterize the yielding/ buckling of the cylindrical pressure vessel. The fuzzy decision maker was used to estimate the thickness of the steel liner and the number of composite layers. The ratio of stresses on the composite fibers and the working pressure as well as the ratio of stresses on the composite fibers and the burst (failure) pressure were assessed. ANSYS nonlinear finite element solver was used to analyze the residual stress in the steel liner induced due to an auto-frettage process. Result of analysis verified that carbon fibers are the most suitable reinforcement to increase strength of cylinder while the weight stayed appreciably low.

  13. Friction Stir Welding of Steel Alloys

    Science.gov (United States)

    Ding, R. Jeffrey; Munafo, Paul M. (Technical Monitor)

    2001-01-01

    The friction stir welding process has been developed primarily for the welding of aluminum alloys. Other higher melting allows such, as steels are much more difficult to join. Special attention must be given to pin tool material selection and welding techniques. This paper addresses the joining of steels and other high melting point materials using the friction stir welding process. Pin tool material and welding parameters will be presented. Mechanical properties of weldments will also be presented. Significance: There are many applications for the friction stir welding process other than low melting aluminum alloys. The FSW process can be expanded for use with high melting alloys in the pressure vessel, railroad and ship building industries.

  14. [Large vessel vasculitides].

    Science.gov (United States)

    Morović-Vergles, Jadranka; Puksić, Silva; Gracanin, Ana Gudelj

    2013-01-01

    Large vessel vasculitis includes Giant cell arteritis and Takayasu arteritis. Giant cell arteritis is the most common form of vasculitis affect patients aged 50 years or over. The diagnosis should be considered in older patients who present with new onset of headache, visual disturbance, polymyalgia rheumatica and/or fever unknown cause. Glucocorticoides remain the cornerstone of therapy. Takayasu arteritis is a chronic panarteritis of the aorta ant its major branches presenting commonly in young ages. Although all large arteries can be affected, the aorta, subclavian and carotid arteries are most commonly involved. The most common symptoms included upper extremity claudication, hypertension, pain over the carotid arteries (carotidynia), dizziness and visual disturbances. Early diagnosis and treatment has improved the outcome in patients with TA.

  15. Vessel segmentation in screening mammograms

    Science.gov (United States)

    Mordang, J. J.; Karssemeijer, N.

    2015-03-01

    Blood vessels are a major cause of false positives in computer aided detection systems for the detection of breast cancer. Therefore, the purpose of this study is to construct a framework for the segmentation of blood vessels in screening mammograms. The proposed framework is based on supervised learning using a cascade classifier. This cascade classifier consists of several stages where in each stage a GentleBoost classifier is trained on Haar-like features. A total of 30 cases were included in this study. In each image, vessel pixels were annotated by selecting pixels on the centerline of the vessel, control samples were taken by annotating a region without any visible vascular structures. This resulted in a total of 31,000 pixels marked as vascular and over 4 million control pixels. After training, the classifier assigns a vesselness likelihood to the pixels. The proposed framework was compared to three other vessel enhancing methods, i) a vesselness filter, ii) a gaussian derivative filter, and iii) a tubeness filter. The methods were compared in terms of area under the receiver operating characteristics curves, the Az values. The Az value of the cascade approach is 0:85. This is superior to the vesselness, Gaussian, and tubeness methods, with Az values of 0:77, 0:81, and 0:78, respectively. From these results, it can be concluded that our proposed framework is a promising method for the detection of vessels in screening mammograms.

  16. Americium behaviour in plastic vessels

    Energy Technology Data Exchange (ETDEWEB)

    Legarda, F.; Herranz, M. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Idoeta, R., E-mail: raquel.idoeta@ehu.e [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Abelairas, A. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain)

    2010-07-15

    The adsorption of {sup 241}Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of {sup 241}Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of {sup 241}Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

  17. The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn; Lu, Feng; Wang, Rongshan; Huang, Ping; Liu, Xiangbin; Zhang, Guodong; Xu, Chaoliang

    2015-07-15

    Highlights: • The conservative and non-conservative assumptions in the codes were shown. • The influence of different loads on the SM was given. • The unloading effect of the cladding was studied. • A concentrated reflection of the safety was shown based on 3-D FE analyses. - Abstract: The deterministic structural integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. While the nil-ductility-transition temperature (RT{sub NDT}) parameter is widely used, the influence of fluence and temperature distributions along the thickness of the base metal wall cannot be reflected in the comparative analysis. This paper introduces the method using a structure safety margin (SM) parameter which is based on a comparison between the material toughness (the fracture initiation toughness K{sub IC} or fracture arrest toughness K{sub Ia}) and the stress intensity factor (SIF) along the crack front for the integrity analysis of a RPV subjected to PTS transients. A 3-D finite element model is used to perform fracture mechanics analyses considering both crack initiation assessment and arrest assessment. The results show that the critical part along the crack front is always the clad-base metal interface point (IP) rather than the deepest point (DP) for either crack initiation assessment or crack arrest assessment under the thermal load. It is shown that the requirement in Regulatory Guide 1.154 that ‘axial flaws with depths less than 20% of the wall thickness and all circumferential flaws should be modeled in infinite length’ may be non-conservative. As the assessment result is often poor universal for a given material, crack and transient, caution is recommended in the safety assessment, especially for the IP. The SIF reduces under the thermal or pressure load if the map cracking (MC) effect is considered. Therefore, the assumption in the ASME and RCCM codes that the cladding should be taken into account in

  18. Tool steels. 5. edition

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, G.; Krauss, G.; Kennedy, R.

    1998-12-31

    The revision of this authoritative work contains a significant amount of new information from the past nearly two decades presented in an entirely new outline, making this a must have reference for engineers involved in tool-steel production, as well as in the selection and use of tool steels in metalworking and other materials manufacturing industries. The chapter on tool-steel manufacturing includes new production processes, such as electroslag refining, vacuum arc remelting, spray deposition processes (Osprey and centrifugal spray), and powder metal processing. The seven chapters covering tool-steel types in the 4th Edition have been expanded to 11 chapters covering nine main groups of tool steels as well as other types of ultrahigh strength steels sometimes used for tooling. Each chapter discusses in detail processing, composition, and applications specific to the particular group. In addition, two chapters have been added covering surface modification and trouble shooting production and performance problems.

  19. Heat treatments in a conventional steel to reproduce the microstructure of a nuclear grade steel; Tratamientos termicos en un acero convencional para reproducir la microestructura de un acero grado nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rosalio G, M.

    2014-07-01

    The ferritic steels used in the manufacture of pressurized vessels of Boiling Water Reactors (BWR) suffer degradation in their mechanical properties due to damage caused by the neutron fluxes of high energy bigger to a Mega electron volt (E> 1 MeV) generated in the reactor core. The materials with which the pressurized vessels of nuclear reactors cooled by light water are built correspond to low alloy ferritic steels. The effect of neutron irradiation on these steels is manifested as an increase in hardness, mechanical strength, with the consequent decrease in ductility, fracture toughness and an increase in temperature of ductile-brittle transition. The life of a BWR is 40 years, its design must be considered sufficient margin of safety because pressure forces experienced during operation, maintenance and testing of postulated accident conditions. It is necessary that under these conditions the vessel to behave ductile and likely to propagate a fracture is minimized. The vessels of light water nuclear reactors have a bainite microstructure. Specifically, the reactor vessels of the nuclear power plant of Laguna Verde (Veracruz, Mexico) are made of a steel Astm A-533, Grade B Class 1. At present they are carrying out some welding tests for the construction of a model of a BWR, however, to use nuclear grade steel such as Astm A-533 to carry out some of the welding tests, is very expensive; perform these in a conventional material provides basic information. Although the microstructure present in the conventional material does not correspond exactly to the degree of nuclear material, it can take of reference. Therefore, it is proposed to conduct a pilot study to establish the thermal treatment that reproduces the microstructure of nuclear grade steel, in conventional steel. The resulting properties of the conventional steel samples will be compared to a JRQ steel, that is a steel Astm A-533, Grade B Class 1, provided by IAEA. (Author)

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

    Science.gov (United States)

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

    2014-10-28

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

  1. Ultrasonic guided wave detection of scatterers on large clad steel plates

    Science.gov (United States)

    Gong, Peng; Harley, Joel B.; Berges, Mario; Junker, Warren R.; Greve, David W.; Oppenheim, Irving J.

    2016-04-01

    "Clad steel" refers to a thick carbon steel structural plate bonded to a corrosion resistant alloy (CRA) plate, such as stainless steel or titanium, and is widely used in industry to construct pressure vessels. The CRA resists the chemically aggressive environment on the interior, but cannot prevent the development of corrosion losses and cracks that limit the continued safe operation of such vessels. At present there are no practical methods to detect such defects from the exposed outer surface of the thick carbon steel plate, often necessitating removing such vessels from service and inspecting them visually from the interior. In previous research, sponsored by industry to detect and localize damage in pressurized piping systems under operational and environmental changes, we investigated a number of data-driven signal processing methods to extract damage information from ultrasonic guided wave pitch-catch records. We now apply those methods to relatively large clad steel plate specimens. We study a sparse array of wafer-type ultrasonic transducers adhered to the carbon steel surface, attempting to localize mass scatterers grease-coupled to the stainless steel surface. We discuss conditions under which localization is achieved by relatively simple first-arrival methods, and other conditions for which data-driven methods are needed; we also discuss observations of plate-like mode properties implied by these results.

  2. BORONIZING OF STEEL

    OpenAIRE

    ULUKÖY, Arzum; CAN, Ahmet Çetin

    2006-01-01

    Boride layer has many advantages in comparison with traditional hardening methods. The boride layer has high hardening value and keeps it's hardeness at high temperatures, and it also shows favorible properties, such as the resistance to wear, oxidation and corrosion. The process can be applied at variety of materials, for instance steel, cast iron, cast steel, nickel and cobalt alloys and cermets. In this rewiew, boronizing process properties, boride layer on steel surfaces and specification...

  3. Development of a New Class of Fe-3Cr-W(V) Ferritic Steels for Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Jawad, Mann; Sikka, Vinod K.

    2005-04-06

    The project described in this report dealt with improving the materials performance and fabrication for hydrotreating reactor vessels, heat recovery systems, and other components for the petroleum and chemical industries. These reactor vessels can approach ship weights of about 300 tons with vessel wall thicknesses of 3 to 8 inches. They are typically fabricated from Fe-Cr-Mo alloy steels, containing 1.25 to 12% chromium and 1 to 2% molybdenum. The goal of this project was to develop Fe-Cr-W(V) steels that can perform similar duties, in terms of strength at high temperatures, but will weigh less and thereby save energy.

  4. 50 CFR 648.8 - Vessel identification.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 8 2010-10-01 2010-10-01 false Vessel identification. 648.8 Section 648.8... identification. (a) Vessel name and official number. Each fishing vessel subject to this part and over 25 ft (7.6... or ocean quahog vessels licensed under New Jersey law may use the appropriate vessel identification...

  5. Modern Steel Framed Schools.

    Science.gov (United States)

    American Inst. of Steel Construction, Inc., New York, NY.

    In view of the cost of structural framing for school buildings, ten steel-framed schools are examined to review the economical advantages of steel for school construction. These schools do not resemble each other in size, shape, arrangement or unit cost; some are original in concept and architecture, and others are conservative. Cost and…

  6. Fluid flow separation in a reactor pressure vessel during an ECC injection. Single phase flow and two phase flow (air-water) experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Thierry Bichet; Alain Martin [EDF - Research and Development Division - Fluid Mechanics and Heat Transfert 6, quai Watier - B.P. 49 - 78401 Chatou CEDEX 01 (France); Frederic Beaud [EDF/ Industry - Basic Design Department., 12-14, Avenue Dutrievoz 69628 Villeurbanne CEDEX (France)

    2005-07-01

    Full text of publication follows: Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) integrity has been performed. A simplified analysis has shown that the most severe loading conditions are given by the small break loss of coolant accidents due to the pressurized injection of cold water (9 deg. C) into the cold leg and down comer of the RPV. During these transient scenarios, single or two-phase (uncovered cold leg) flows have been shown in the cold leg, depending on the crack size and RPV model (900 MWe or 1300 MWe). An experimental study has been carried out, on the one hand, to consolidate the numerical results obtained with CFD home code (Code-Saturne) which mainly showed the stratified flow in the cold leg and the fluid flow separation and its oscillations in the down comer during a single phase scenario. These physical phenomena are important for the thermal RPV loading assessment. On the other hand, the absence of experimental two-phase data necessitated to carry out an experimental study around the mixing area behavior (free surface, stratified flow) during an ECC injection with an uncovered cold leg. The new EDF R and D mock up, called HYBISCUS, is a facility which is made out of Plexiglas (atmosphere pressure) and represents a half scale CP0 geometry with one cold leg and part of the down comer. The mock up modularity allows us to insert representative ECC nozzles and a thermal shield. In reference to the reactor scenarios, the experimental operating conditions are derived from the conservation of the density effects (Froude number). For that, a heated salted water flow is used to represent the ECC injection whereas water represents the cold leg fluid. This mock up has been defined in order to represent single phase flow (cold leg and down comer full of water) or two-phase flow (uncovered cold leg) ECC scenarios. This paper reports experimental results

  7. 2D Fast Vessel Visualization Using a Vessel Wall Mask Guiding Fine Vessel Detection

    Directory of Open Access Journals (Sweden)

    Sotirios Raptis

    2010-01-01

    and then try to approach the ridges and branches of the vasculature's using fine detection. Fine vessel screening looks into local structural inconsistencies in vessels properties, into noise, or into not expected intensity variations observed inside pre-known vessel-body areas. The vessels are first modelled sufficiently but not precisely by their walls with a tubular model-structure that is the result of an initial segmentation. This provides a chart of likely Vessel Wall Pixels (VWPs yielding a form of a likelihood vessel map mainly based on gradient filter's intensity and spatial arrangement parameters (e.g., linear consistency. Specific vessel parameters (centerline, width, location, fall-away rate, main orientation are post-computed by convolving the image with a set of pre-tuned spatial filters called Matched Filters (MFs. These are easily computed as Gaussian-like 2D forms that use a limited range sub-optimal parameters adjusted to the dominant vessel characteristics obtained by Spatial Grey Level Difference statistics limiting the range of search into vessel widths of 16, 32, and 64 pixels. Sparse pixels are effectively eliminated by applying a limited range Hough Transform (HT or region growing. Major benefits are limiting the range of parameters, reducing the search-space for post-convolution to only masked regions, representing almost 2% of the 2D volume, good speed versus accuracy/time trade-off. Results show the potentials of our approach in terms of time for detection ROC analysis and accuracy of vessel pixel (VP detection.

  8. Southeast Region Headboat Survey-Vessel list/Vessel Directory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This is a database of vessels that have been on the SRHS through time, their owners/operators, marinas/docks and their contact information. This assists in...

  9. Optimal control of the vessel motion process at the broadside slip

    Directory of Open Access Journals (Sweden)

    Anton A. Omelchuk

    2015-12-01

    Full Text Available The main purpose during the mechanized launching or lifting of a vessel using broadside slip is the uniform velocity motion providing of the object (vessel with launching trolleys and the guaranteed prevention of emergencies. Aim: The aim of this work is the formulation and optimal control solution of a slip type vessel-lifting complex for providing the uniform velocity, coordinated, gradual relocation of the large vessels. Materials and Methods: The paper studies the matters of real-time control optimization of vessel-lifting complex during launching or lifting the vessel. Results: It is proposed to control the relocation of large object by funicular forces matching of the steel wire ropes of different motorized drives with account of current situation. The model in state space is obtained. This model describes the vessel relocation during the launching/lifting process. The optimal control solution of motorized multidrive system with goal functional is formulated. This solution allows minimizing the divergence of movement parameter values from given ones. It is substantiated the advisability of the use of adaptive control methods with observer to provide the corresponding reliability of vessel-lifting complex functioning.

  10. 2013 East Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  11. SC/OQ Vessel Database

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Data tables holding information for the Surf Clam/Ocean Quahog vessel and dealer/processor logbooks (negative and positive), as well as individual tag information...

  12. 2011 Great Lakes Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  13. 2011 West Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  14. 2013 Great Lakes Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  15. 2011 East Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  16. Integrin binding: Sticking around vessels

    Science.gov (United States)

    Blatchley, Michael R.; Gerecht, Sharon

    2017-09-01

    A study demonstrates that controlled integrin binding on a biomaterial was capable of promoting vascular cell sprouting and formation of a non-leaky blood vessel network in a healthy and diseased state.

  17. Transposition of the great vessels

    Science.gov (United States)

    ... vessel called the ductus arteriosus open, allowing some mixing of the 2 blood circulations. A procedure using ... they are not already immune. Eating well, avoiding alcohol, and controlling diabetes both before and during pregnancy ...

  18. 2013 West Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  19. Vessel Permit System Data Set

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — GARFO issues federal fishing permits annually to owners of fishing vessels who fish in the Greater Atlantic region, as required by federal regulation. These permits...

  20. 2011 Tug Towing Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  1. Caribbean PR Logbook Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch (landed catch) and effort for fishing trips made by vessels fishing in Puerto Rico. The catch and effort data for the entire trip are...

  2. Coastal Discard Logbook Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data on the type and amount of marine resources that are discarded or interacted with by vessels that are selected to report to the Southeast...

  3. RPV steam generator pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Strosnider, J.

    1996-03-01

    As the types of SG tube degradation affecting PWR SGs has changed, and improvements in tube inspection and repair technology have occurred, current SG regulatory requirements and guidance have become increasingly out of date. This regulatory situation has been dealt with on a plant-specific basis, however to resolve this problem in the long term, the NRC has begun development of a performance-based rule. As currently structured, the proposed steam generator rule would require licensees to implement SG programs that monitor the condition of the steam generator tubes against accepted performance criteria to provide reasonable assurance that the steam generator tubes remain capable of performing their intended safety functions. Currently the staff is developing three performance criteria that will ensure the tubes can continue to perform their safety function and therefore satisfy the SG rule requirements. The staff, in developing the criteria, is striving to ensure that the performance criteria have the two key attributes of being (1) measurable (enabling the tube condition to be {open_quotes}measured{close_quotes} against the criteria) and (2) tolerable (ensuring that failures to meet the criteria do not result in unacceptable consequences). A general description of the criteria are: (1) Structural integrity criteria: Ensures that the structural integrity of the SG tubes is maintained for the operating cycle consistent with the margins intended by the ASME Code. (2) Leakage integrity criteria: Ensures that postulated accident leakages and the associated dose releases are limited relative to 10 CFR Part 50 guidelines and 10 CFR Part 50 Appendix A GDC 19. (3) Operational leakage criteria: Ensures that the operating unit will be shut down as a defense-in depth measure when operational SG tube leakage exceeds established leakage limits.

  4. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  5. Prosopomorphic vessels from Moesia Superior

    Directory of Open Access Journals (Sweden)

    Nikolić Snežana

    2008-01-01

    Full Text Available The prosopomorphic vessels from Moesia Superior had the form of beakers varying in outline but similar in size. They were wheel-thrown, mould-made or manufactured by using a combination of wheel-throwing and mould-made appliqués. Given that face vessels are considerably scarcer than other kinds of pottery, more than fifty finds from Moesia Superior make an enviable collection. In this and other provinces face vessels have been recovered from military camps, civilian settlements and necropolises, which suggests that they served more than one purpose. It is generally accepted that the faces-masks gave a protective role to the vessels, be it to protect the deceased or the family, their house and possessions. More than forty of all known finds from Moesia Superior come from Viminacium, a half of that number from necropolises. Although tangible evidence is lacking, there must have been several local workshops producing face vessels. The number and technological characteristics of the discovered vessels suggest that one of the workshops is likely to have been at Viminacium, an important pottery-making centre in the second and third centuries.

  6. BORONIZING OF STEEL

    Directory of Open Access Journals (Sweden)

    Arzum ULUKÖY

    2006-02-01

    Full Text Available Boride layer has many advantages in comparison with traditional hardening methods. The boride layer has high hardening value and keeps it's hardeness at high temperatures, and it also shows favorible properties, such as the resistance to wear, oxidation and corrosion. The process can be applied at variety of materials, for instance steel, cast iron, cast steel, nickel and cobalt alloys and cermets. In this rewiew, boronizing process properties, boride layer on steel surfaces and specifications and the factors that effect boride layer are examined

  7. Ex-vessel core catcher materials interactions. Annual progress report. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, D.G.; Zehms, E.H.; Ang, C.Y.; McClelland, J.D.; Meyer, R.A.; vanPaassen, H.L.L.

    1976-10-30

    A twelve-month program to investigate ex-vessel core catcher materials interactions has been completed. The investigations, involving depleted uranium dioxide, magnesia brick, stainless steel, and low-carbon steel, were conducted in furnaces and associated facilities existing at Aerospace, which were modified to process molten and solidified radioactive samples. In addition to developing efficient methods for the melting, pouring, and sustained heating of UO/sub 2/, extensive sample characterizations and microanalyses were performed. Theoretical analyses were also made in data interpretation for the purpose of understanding the interaction kinetics.

  8. Swedish Work on Brittle-Fracture Problems in Nuclear Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M.

    1966-03-15

    After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO

  9. Glass Stronger than Steel

    Science.gov (United States)

    Yarris, Lynn

    2011-03-28

    A new type of damage-tolerant metallic glass, demonstrating a strength and toughness beyond that of steel or any other known material, has been developed and tested by a collaboration of researchers from Berkeley Lab and Caltech.

  10. Metallurgy: Printing steels

    Science.gov (United States)

    Todd, Iain

    2018-01-01

    Additive manufacturing has been used to fabricate a common stainless steel, which imparts a unique microstructure to this material, making it stronger and more ductile than that produced with conventional methods.

  11. Automated method for identification and artery-venous classification of vessel trees in retinal vessel networks.

    Directory of Open Access Journals (Sweden)

    Vinayak S Joshi

    Full Text Available The separation of the retinal vessel network into distinct arterial and venous vessel trees is of high interest. We propose an automated method for identification and separation of retinal vessel trees in a retinal color image by converting a vessel segmentation image into a vessel segment map and identifying the individual vessel trees by graph search. Orientation, width, and intensity of each vessel segment are utilized to find the optimal graph of vessel segments. The separated vessel trees are labeled as primary vessel or branches. We utilize the separated vessel trees for arterial-venous (AV classification, based on the color properties of the vessels in each tree graph. We applied our approach to a dataset of 50 fundus images from 50 subjects. The proposed method resulted in an accuracy of 91.44% correctly classified vessel pixels as either artery or vein. The accuracy of correctly classified major vessel segments was 96.42%.

  12. Life after Steel

    Science.gov (United States)

    Mangan, Katherine

    2013-01-01

    Bobby Curran grew up in a working-class neighborhood in Baltimore, finished high school, and followed his grandfather's steel-toed bootprints straight to Sparrows Point, a 3,000-acre sprawl of industry on the Chesapeake Bay. College was not part of the plan. A gritty but well-paying job at the RG Steel plant was Mr. Curran's ticket to a secure…

  13. Effect of Heat Treatment on Microstructure and Hardness of Grade 91 Steel

    Directory of Open Access Journals (Sweden)

    Triratna Shrestha

    2015-01-01

    Full Text Available Grade 91 steel (modified 9Cr-1Mo steel is considered a prospective material for the Next Generation Nuclear Power Plant for application in reactor pressure vessels at temperatures of up to 650 °C. In this study, heat treatment of Grade 91 steel was performed by normalizing and tempering the steel at various temperatures for different periods of time. Optical microscopy, scanning and transmission electron microscopy in conjunction with microhardness profiles and calorimetric plots were used to understand the microstructural evolution including precipitate structures and were correlated with mechanical behavior of the steel. Thermo-Calc™ calculations were used to support the experimental work. Furthermore, carbon isopleth and temperature dependencies of the volume fraction of different precipitates were constructed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Broadhead, B.L. [Oak Ridge National Lab., TN (United States); Suzuki, M.; Kohsaka, A. [Japan Atomic Energy Research Institute, Tokai (Japan)

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  15. 50 CFR 697.8 - Vessel identification.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Vessel identification. 697.8 Section 697.8 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND ATMOSPHERIC ADMINISTRATION....8 Vessel identification. (a) Vessel name and official number. Each fishing vessel issued a limited...

  16. Mechanosensing in developing lymphatic vessels.

    Science.gov (United States)

    Planas-Paz, Lara; Lammert, Eckhard

    2014-01-01

    The lymphatic vasculature is responsible for fluid homeostasis, transport of immune cells, inflammatory molecules, and dietary lipids. It is composed of a network of lymphatic capillaries that drain into collecting lymphatic vessels and ultimately bring fluid back to the blood circulation. Lymphatic endothelial cells (LECs) that line lymphatic capillaries present loose overlapping intercellular junctions and anchoring filaments that support fluid drainage. When interstitial fluid accumulates within tissues, the extracellular matrix (ECM) swells and pulls the anchoring filaments. This results in opening of the LEC junctions and permits interstitial fluid uptake. The absorbed fluid is then transported within collecting lymphatic vessels, which exhibit intraluminal valves that prevent lymph backflow and smooth muscle cells that sequentially contract to propel lymph.Mechanotransduction involves translation of mechanical stimuli into biological responses. LECs have been shown to sense and respond to changes in ECM stiffness, fluid pressure-induced cell stretch, and fluid flow-induced shear stress. How these signals influence LEC function and lymphatic vessel growth can be investigated by using different mechanotransduction assays in vitro and to some extent in vivo.In this chapter, we will focus on the mechanical forces that regulate lymphatic vessel expansion during embryonic development and possibly secondary lymphedema. In mouse embryos, it has been recently shown that the amount of interstitial fluid determines the extent of lymphatic vessel expansion via a mechanosensory complex formed by β1 integrin and vascular endothelial growth factor receptor-3 (VEGFR3). This model might as well apply to secondary lymphedema.

  17. Test of 6-in. -thick pressure vessels. Series 4: intermediate test vessels V-5 and V-9 with inside nozzle corner cracks. [BWR and PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-08-01

    Failure testing is described for two 99-cm-diam (39-in.), 15.2-cm-thick (6-in.) steel pressure vessels, each containing one flawed nozzle. Vessel V-5 was tested at 88/sup 0/C (190/sup 0/F) and failed by leaking without fracturing after extensive stable crack growth. Vessel V-9 was tested at 25/sup 0/C (75/sup 0/F) and failed by fracturing. Material properties measured before the tests were used for pretest and posttest fracture analyses. Test results supported by analysis indicate that inside nozzle corner cracks are not subject to plane strain under pressure loading. The preparation of inside nozzle corner cracks is described in detail. Extensive experimental data are tabulated and plotted.

  18. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  19. Grounding Damage to Conventional Vessels

    DEFF Research Database (Denmark)

    Lützen, Marie; Simonsen, Bo Cerup

    2003-01-01

    regulations for design of bottom compartment layout with regard to grounding damages are largely based on statistical damage data. New and updated damage statistics holding 930 grounding accident records has been investigated. The bottom damage statistics is compared to current regulations for the bottom......The present paper is concerned with rational design of conventional vessels with regard to bottom damage generated in grounding accidents. The aim of the work described here is to improve the design basis, primarily through analysis of new statistical data for grounding damage. The current...... for the relation between the amount of deformed structure and the energy absorption. Finally, the paper shows how damage statistics for existing, conventional vessels can be used together with theoretical prediction methods for determining grounding damage distributions for new vessel types not included...

  20. 19 CFR 4.5 - Government vessels.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 1 2010-04-01 2010-04-01 false Government vessels. 4.5 Section 4.5 Customs Duties... VESSELS IN FOREIGN AND DOMESTIC TRADES Arrival and Entry of Vessels § 4.5 Government vessels. (a) No... that is the property of, the U.S. Department of Defense (DoD) will be treated as a Government vessel...

  1. ATLAS Supplier Award for the ECT Vacuum Vessels

    CERN Multimedia

    Jenni, P

    On 12 February the Netherlands firm Schelde Exotech was awarded the ATLAS Supplier Award for the construction of the two vacuum vessels for the ATLAS End- Cap Toroid (ECT) magnets. ATLAS Supplier Award ceremonies have now become something of a tradition. For the third consecutive year, ATLAS has given best supplier awards for the most exceptional contributors to the construction of the detector. The Netherlands firm Schelde Exotech has just received the award for the construction of the two vacuum vessels for the ECTs. With a diameter of 11 metres and a volume of 550 cubic metres, the ECT vacuum vessels are obviously impressive in scale. They consist of large aluminium plates and a stainless steel central bore tube. In order to obtain the required undulations, the firm had to develop a special assembly and welding technique. Despite the chambers' imposing size, a very high degree of precision has been achieved in their geometry. Moreover, the chambers, which were delivered in July 2002 to CERN, were built i...

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

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J

    1983-01-01

    Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development.

  3. [Pulmonary blood vessels in goats].

    Science.gov (United States)

    Roos, H; Hegner, K; Vollmerhaus, B

    1999-05-01

    The blood vessels in the lung of the goat, which until now have received little attention, are described in detail for the first time. With regard to the segments of the lung, blood vessels are bronchovascular units in the lobi craniales, lobus medius and lobus accessorius, but bronchoartery units in the lobi caudales. We investigated the types of branches of the Aa. pulmonales dextra et sinistra, the inter- and intraspecific principles of the outlet of the pulmonary veins and the importance of bronchopulmonary segmentation of the lungs.

  4. Steady-state CFD simulations of an EPR™ reactor pressure vessel: A validation study based on the JULIETTE experiments

    Energy Technology Data Exchange (ETDEWEB)

    Puragliesi, R., E-mail: riccardo.puragliesi@psi.ch [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland); Zhou, L. [Science and Technology on Reactor System Design Technology Laboratory, NPIC, Chengdu (China); Zerkak, O.; Pautz, A. [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland)

    2016-04-15

    Highlights: • CFD validation of k–ε (RANS model of EPR RPV. • Flat inlet velocity profile is not sufficient to correctly predict the pressure drops. • Swirl is responsible for asymmetric loads at the core barrel. • Parametric study to the turbulent Schmidt number for better predictions of passive-scalar transport. • The optimal turbulent Schmidt number was found to be one order of magnitude smaller than the standard value. - Abstract: Validating computational fluid dynamics (CFD) models against experimental measurements is a fundamental step towards a broader acceptance of CFD as a tool for reactor safety analysis when best-estimate one-dimensional thermal-hydraulic codes present strong modelling limitations. In the present paper numerical results of steady-state RANS analyses are compared to pressure, volumetric flow rate and concentration distribution measurements in different locations of an Areva EPR™ reactor pressure vessel (RPV) mock-up named JULIETTE. Several flow configurations are considered: Three different total volumetric flow rates, cold leg velocity field with or without swirl, three or four reactor coolant pumps functioning. Investigations on the influence of two types of inlet boundary profiles (i.e. flat or 1/7th power-law) and the turbulent Schmidt number have shown that the first affects sensibly the pressure loads at the core barrel whereas the latter parameter strongly affects the transport and the mixing of the tracer (passive scalar) and consequently its distribution at the core inlet. Furthermore, the introduction of an integral parameter as the swirl number has helped to decrease the large epistemic uncertainty associated with the swirling device. The swirl is found to be the cause of asymmetric loads on the walls of the core barrel and also asymmetries are enhanced for the tracer concentration distribution at the core inlet. The k–ϵ CFD model developed with the commercial code STAR-CCM+ proves to be able to predict

  5. Recertification of the air and methane storage vessels at the Langley 8-foot high-temperature structures tunnel

    Science.gov (United States)

    Hudson, C. M.; Girouard, R. L.; Young, C. P., Jr.; Petley, D. H.; Hudson, J. L., Jr.; Hudgins, J. L.

    1977-01-01

    This center operates a number of sophisticated wind tunnels in order to fulfill the needs of its researchers. Compressed air, which is kept in steel storage vessels, is used to power many of these tunnels. Some of these vessels have been in use for many years, and Langley is currently recertifying these vessels to insure their continued structural integrity. One of the first facilities to be recertified under this program was the Langley 8-foot high-temperature structures tunnel. This recertification involved (1) modification, hydrotesting, and inspection of the vessels; (2) repair of all relevant defects; (3) comparison of the original design of the vessel with the current design criteria of Section 8, Division 2, of the 1974 ASME Boiler and Pressure Vessel Code; (4) fracture-mechanics, thermal, and wind-induced vibration analyses of the vessels; and (5) development of operating envelopes and a future inspection plan for the vessels. Following these modifications, analyses, and tests, the vessels were recertified for operation at full design pressure (41.4 MPa (6000 psi)) within the operating envelope developed.

  6. Recent advances in creep-resistant steels for power plant applications

    Indian Academy of Sciences (India)

    M. Senthilkumar (Newgen Imaging) 1461 1996 Oct 15 13:05:22

    Interaction of Steels with Hydrogen in Petroleum Industry. Pressure Vessel and Pipeline Service (New York: Mater. Properties Council) vol. 2, pp 607–618. Czyrska-Filemonowicz A, Penkalla H J, Zielinska-Lipiec A, Ennis P J 2001 Proc. 9th Int. Conf. on. Creep Resistant Metallic Materials, Prague, (ed.) J Purmensky, pp 204– ...

  7. 76 FR 77964 - High Pressure Steel Cylinders From the People's Republic of China: Preliminary Determination of...

    Science.gov (United States)

    2011-12-15

    ... specifications and permanently impressed with ISO or UN symbols. Also excluded from the investigation are... ventures between Chinese and foreign companies, or are wholly Chinese-owned companies, the Department must... Chengyu Co., Ltd.; and Zhuolu High Pressure Vessel Co., Ltd. \\63\\ See, e.g., Prestressed Concrete Steel...

  8. Research and Service Experience with Environmentally-Assisted Cracking in Carbon and Low-Alloy Steels in High-Temperature Water

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, Hans-Peter; Ritter, Stefan [Paul Scherrer Inst., Laboratory for Materials Behaviour, Villigen (Switzerland). Nuclear Energy and Safety Research Dept.

    2005-11-15

    , particularly in BWR service, most often in piping and, rarely in the reactor pressure vessel (RPV) itself. Oxidizing conditions, usually dissolved oxygen (DO), and relevant dynamic straining were always involved. These cases were either attributed to strain-induced corrosion cracking (SICC) or corrosion fatigue (CF) and could be readily rationalized by the experimental background knowledge. Both service experience and experimental/mechanistic background knowledge confirm the high resistance of C and LAS to SCC under stationary power operation and static loading conditions and clearly reveal, that slow, positive (tensile) straining, with associated plastic yielding and sufficiently oxidizing conditions are essential for EAC initiation in HT water. Based on the experimental/mechanistic background knowledge and service experience different remedial and mitigation actions have been qualified and successfully applied, which further reduced the low EAC cracking frequency in the field. In spite of the absence of SCC in the field, several unfavourable critical parameter combinations, which can lead to sustained, fast SCC with crack growth rates (CGRs) well above the BWRVIP-60 SCC DLs have been identified. Many of them appear atypical for current BWR plant operation with properly manufactured C and LAS components, but some might occur during service, at least temporarily under faulted conditions or in components with fabrication deficiencies. Although there are open questions and potentials for improvements in all fields, from a safety perspective, the special emphasis of research should be placed to these conditions, and in particular, to an improved identification/quantification of the boundaries/thresholds for the transition from low to high/accelerated SCC CGRs. In this context, research should be focused on the effects of chloride transients and dynamic strain ageing (DSA)/yield stress (YS) on the SCC crack growth behaviour of C and LAS and of weld heat-affected zone (HAZ

  9. Commercial Passenger Fishing Vessel Fishery

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains the logbook data from U.S.A. Commercial Passenger Fishing Vessels (CPFV) fishing in the U.S.A. EEZ and in waters off of Baja California, from...

  10. Pressure vessel and method therefor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. BPC 157 and blood vessels.

    Science.gov (United States)

    Seiwerth, Sven; Brcic, Luka; Vuletic, Lovorka Batelja; Kolenc, Danijela; Aralica, Gorana; Misic, Marija; Zenko, Anita; Drmic, Domagoj; Rucman, Rudolf; Sikiric, Predrag

    2014-01-01

    This review focuses on the described effects of BPC 157 on blood vessels after different types of damage, and elucidate by investigating different aspects of vascular response to injury (endothelium damage, clotting, thrombosis, vasoconstriction, vasodilatation, vasculoneogenesis and edema formation) especially in connection to the healing processes. In this respect, BPC 157 was concluded to be the most potent angiomodulatory agent, acting through different vasoactive pathways and systems (e.g. NO, VEGF, FAK) and leading to optimization of the vascular response followed, as it has to be expected, by optimization of the healing process. Formation of new blood vessels involves two main, partly overlapping mechanisms, angiogenesis and vasculogenesis. The additional mechanism of arteriogenesis is involved in the formation of collaterals. In conjunction with blood vessel function, we at least have to consider leakage of fluid/proteins/plasma, resulting in edema/exudate formation as well as thrombogenesis. Blood vessels are also strongly involved in tumor biology. In this aspect, we have neoangiogenesis resulting in pathological vascularization, vascular invasion resulting in release of metastatic cells and the phenomenon of homing resulting in formation of secondary tumors--metastases.

  12. The determinants of fishing vessel accident severity.

    Science.gov (United States)

    Jin, Di

    2014-05-01

    The study examines the determinants of fishing vessel accident severity in the Northeastern United States using vessel accident data from the U.S. Coast Guard for 2001-2008. Vessel damage and crew injury severity equations were estimated separately utilizing the ordered probit model. The results suggest that fishing vessel accident severity is significantly affected by several types of accidents. Vessel damage severity is positively associated with loss of stability, sinking, daytime wind speed, vessel age, and distance to shore. Vessel damage severity is negatively associated with vessel size and daytime sea level pressure. Crew injury severity is also positively related to the loss of vessel stability and sinking. Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Thermochemical surface engineering of steels

    DEFF Research Database (Denmark)

    Thermochemical Surface Engineering of Steels provides a comprehensive scientific overview of the principles and different techniques involved in thermochemical surface engineering, including thermodynamics, kinetics principles, process technologies and techniques for enhanced performance of steels...

  14. Brazing titanium to stainless steel

    Science.gov (United States)

    Batista, R. I.

    1980-01-01

    Titanium and stainless-steel members are usually joined mechanically for lack of any other effective method. New approach using different brazing alloy and plating steel member with nickel resolves problem. Process must be carried out in inert atmosphere.

  15. A tale of Wootz steel

    National Research Council Canada - National Science Library

    Ranganathan, S; Srinivasan, Sharada

    2006-01-01

    The extraordinary romance and thrilling adventure associated with the tale of wootz steel shows how Indian metallurgists were the world leaders in antiquity in the manufacture of this legendary high-grade steel...

  16. Fatigue damage of steel components

    DEFF Research Database (Denmark)

    Fæster, Søren; Zhang, Xiaodan; Huang, Xiaoxu

    2014-01-01

    Railway rails and the inner ring in roller bearings in wind turbines are both experiencing steel-to-steel contact in small areas with huge loads resulting in extremely high stresses in the base materials......Railway rails and the inner ring in roller bearings in wind turbines are both experiencing steel-to-steel contact in small areas with huge loads resulting in extremely high stresses in the base materials...

  17. 46 CFR 42.05-63 - Ship(s) and vessel(s).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Ship(s) and vessel(s). 42.05-63 Section 42.05-63... BY SEA Definition of Terms Used in This Subchapter § 42.05-63 Ship(s) and vessel(s). The terms ship(s) and vessel(s) are interchangeable or synonymous words, and include every description of watercraft...

  18. Multi Objective Optimization of Weld Parameters of Boiler Steel Using Fuzzy Based Desirability Function

    Directory of Open Access Journals (Sweden)

    M. Satheesh

    2014-01-01

    Full Text Available The high pressure differential across the wall of pressure vessels is potentially dangerous and has caused many fatal accidents in the history of their development and operation. For this reason the structural integrity of weldments is critical to the performance of pressure vessels. In recent years much research has been conducted to the study of variations in welding parameters and consumables on the mechanical properties of pressure vessel steel weldments to optimize weld integrity and ensure pressure vessels are safe. The quality of weld is a very important working aspect for the manufacturing and construction industries. Because of high quality and reliability, Submerged Arc Welding (SAW is one of the chief metal joining processes employed in industry. This paper addresses the application of desirability function approach combined with fuzzy logic analysis to optimize the multiple quality characteristics (bead reinforcement, bead width, bead penetration and dilution of submerged arc welding process parameters of SA 516 Grade 70 steels(boiler steel. Experiments were conducted using Taguchi’s L27 orthogonal array with varying the weld parameters of welding current, arc voltage, welding speed and electrode stickout. By analyzing the response table and response graph of the fuzzy reasoning grade, optimal parameters were obtained. Solutions from this method can be useful for pressure vessel manufacturers and operators to search an optimal solution of welding condition.

  19. Vessel tree extraction using locally optimal paths

    DEFF Research Database (Denmark)

    Lo, Pechin Chien Pau; van Ginneken, Bram; de Bruijne, Marleen

    2010-01-01

    This paper proposes a method to extract vessel trees by continually extending detected branches with locally optimal paths. Our approach uses a cost function from a multi scale vessel enhancement filter. Optimal paths are selected based on rules that take into account the geometric characteristics...... of the vessel tree. Experiments were performed on 10 low dose chest CT scans for which the pulmonary vessel trees were extracted. The proposed method is shown to extract a better connected vessel tree and extract more of the small peripheral vessels in comparison to applying a threshold on the output...

  20. Electrically conductive containment vessel for molten aluminum

    Science.gov (United States)

    Holcombe, C.E.; Scott, D.G.

    1984-06-25

    The present invention is directed to a containment vessel which is particularly useful in melting aluminum. The vessel of the present invention is a multilayered vessel characterized by being electrically conductive, essentially nonwettable by and nonreactive with molten aluminum. The vessel is formed by coating a tantalum substrate of a suitable configuration with a mixture of yttria and particulate metal 10 borides. The yttria in the coating inhibits the wetting of the coating while the boride particulate material provides the electrical conductivity through the vessel. The vessel of the present invention is particularly suitable for use in melting aluminum by ion bombardment.

  1. Thermally Stable Nanocrystalline Steel

    Science.gov (United States)

    Hulme-Smith, Christopher Neil; Ooi, Shgh Woei; Bhadeshia, Harshad K. D. H.

    2017-10-01

    Two novel nanocrystalline steels were designed to withstand elevated temperatures without catastrophic microstructural changes. In the most successful alloy, a large quantity of nickel was added to stabilize austenite and allow a reduction in the carbon content. A 50 kg cast of the novel alloy was produced and used to verify the formation of nanocrystalline bainite. Synchrotron X-ray diffractometry using in situ heating showed that austenite was able to survive more than 1 hour at 773 K (500 °C) and subsequent cooling to ambient temperature. This is the first reported nanocrystalline steel with high-temperature capability.

  2. A-3 steel work completed

    Science.gov (United States)

    2009-01-01

    Stennis Space Center engineers celebrated a key milestone in construction of the A-3 Test Stand on April 9 - completion of structural steel work. Workers with Lafayette (La.) Steel Erector Inc. placed the last structural steel beam atop the stand during a noon ceremony attended by more than 100 workers and guests.

  3. Equipment stainless steel entire versus steels bimetallics clad or overlay; Utilizacao de equipamentos de processo construidos em aco inoxidavel integral versus acos bimetalicos cladeado ou 'overlay'

    Energy Technology Data Exchange (ETDEWEB)

    Batista, Itamar da Silva; Lima, Jadival Carneiro de; Leal, Murilo Fonseca; Cardoso, Amauri dos Santos; Jorjan, Roberto [PETROBRAS S.A., Sao Francisco do Conde, BA (Brazil). Refinaria Landulfo Alves Mataripen (RLAM)

    2008-07-01

    This study does not recommend the use of a pressure vessel made of integral stainless steel, due to the failure mechanisms under stress corrosion assisted by chlorides or polythionic acid. Are presented case studies of literature and analysis of reports of proceedings of RLAM reactors, showing that the materials produced by bimetallic clad overlay or are more appropriate, in terms of integrity, for use in equipment that the internal environment requires austenitic stainless steel specification.

  4. Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Energy Technology

    2002-08-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue {var_epsilon}-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed.

  5. Formation of nano sized ODS clusters in mechanically alloyed NiAl-(Y,Ti,O) alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Deog; Bae Seong Man [KHNP Central Research Institute, Daejeon (Korea, Republic of); Wirth, Brian D. [Department of Nuclear Engineering, Univ. of California, Berkeley (United States)

    2012-10-15

    The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni-Cr-Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn-Mo-Ni low alloy steel were also evaluated.

  6. Microbial-Influenced Corrosion of Corten Steel Compared with Carbon Steel and Stainless Steel in Oily Wastewater by Pseudomonas aeruginosa

    Science.gov (United States)

    Mansouri, Hamidreza; Alavi, Seyed Abolhasan; Fotovat, Meysam

    2015-07-01

    The microbial corrosion behavior of three important steels (carbon steel, stainless steel, and Corten steel) was investigated in semi petroleum medium. This work was done in modified nutrient broth (2 g nutrient broth in 1 L oily wastewater) in the presence of Pseudomonas aeruginosa and mixed culture (as a biotic media) and an abiotic medium for 2 weeks. The behavior of corrosion was analyzed by spectrophotometric and electrochemical methods and at the end was confirmed by scanning electron microscopy. The results show that the degree of corrosion of Corten steel in mixed culture, unlike carbon steel and stainless steel, is less than P. aeruginosa inoculated medium because some bacteria affect Corten steel less than other steels. According to the experiments, carbon steel had less resistance than Corten steel and stainless steel. Furthermore, biofilm inhibits separated particles of those steels to spread to the medium; in other words, particles get trapped between biofilm and steel.

  7. In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2013-01-01

    The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

  8. Guns, Germs and Steel

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 6; Issue 1. Guns, Germs and Steel - A Short History of Everybody for the Last 13,000 years. Suri Venkatachalam. Book Review Volume 6 Issue 1 January 2001 pp 84-88. Fulltext. Click here to view fulltext PDF. Permanent link:

  9. Japan steel mill perspective

    Energy Technology Data Exchange (ETDEWEB)

    Murase, K. [Kobe Steel Ltd., Tokyo (Japan)

    2004-07-01

    The international and Japan's steel industry, the coking coal market, and Japan's expectations from Canada's coal industry are discussed. Japan's steel mills are operating at full capacity. Crude steel production for the first half of 2004 was 55.8 million tons. The steel mills are profitable, but costs are high, and there are difficulties with procuring raw materials. Japan is trying to enhance the quality of coke, in order to achieve higher productivity in the production of pig iron. Economic growth is rising disproportionately in the BRICs (Brazil, Russia, India, and China), with a large increase in coking coal demand from China. On the supply side, there are several projects underway in Australia and Canada to increase production. These include new developments by Elk Valley Coal Corporation, Grande Cache Coal, Western Canadian Coal, and Northern Energy and Mining in Canada. The Elga Mine in the far eastern part of Russia is under development. But the market is expected to remain tight for some time. Japan envisions Canadian coal producers will provide a stable coal supply, expansion of production and infrastructure capabilities, and stabilization of price. 16 slides/overheads are included.

  10. Braze alloy spreading on steel

    Science.gov (United States)

    Siewert, T. A.; Heine, R. W.; Lagally, M. G.

    1978-01-01

    Scanning electron microscopy (SEM) and Auger electron microscopy (AEM) were employed to observe elemental surface decomposition resulting from the brazing of a copper-treated steel. Two types of steel were used for the study, stainless steel (treated with a eutectic silver-copper alloy), and low-carbon steel (treated with pure copper). Attention is given to oxygen partial pressure during the processes; a low enough pressure (8 x 10 to the -5th torr) was found to totally inhibit the spreading of the filler material at a fixed heating cycle. With both types of steel, copper treatment enhanced even spreading at a decreased temperature.

  11. AFSC/FMA/Vessel Assessment Logging

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Vessels fishing trawl gear, vessels fishing hook-and-line and pot gear that are also greater than 57.5 feet overall, and shoreside and floating processing facilities...

  12. Hawaii Abandoned Vessel Inventory, Hawaii Island

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Hawaii Island. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  13. US Virgin Islands Abandoned Vessel Inventory

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for US Virgin Islands. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of...

  14. Hawaii Abandoned Vessel Inventory, Midway Island, NWHI

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Midway Island, NWHI. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of...

  15. Hawaii Abandoned Vessel Inventory, Kure, NWHI

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Kure, NWHI. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  16. Actemra Approved for Certain Blood Vessel Inflammation

    Science.gov (United States)

    ... 165836.html Actemra Approved for Certain Blood Vessel Inflammation Drug will treat adults with a condition called ... to treat adults with giant cell arteritis, an inflammation of the blood vessels (vasculitis). In a media ...

  17. Hawaii Abandoned Vessel Inventory, Maro Reef, NWHI

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Maro Reef, NWHI. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction of coral...

  18. PCs and networking for oceanographic research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Desai, R.G.P.; Desa, E.; Vithayathil, G.

    This paper, first describes briefly the evolution of data acquisition techniques and different system implementation, on board research vessels. A data acquisition system being developed for a coastal research vessel is then described which is based...

  19. Hawaii Abandoned Vessel Inventory, Lisianski Island, NWHI

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NOAA Abandoned Vessel Project Data for Lisianski Island, NWHI. Abandoned vessels pose a significant threat to the NOAA Trust resources through physical destruction...

  20. Purification of Mouse Brain Vessels.

    Science.gov (United States)

    Boulay, Anne-Cécile; Saubaméa, Bruno; Declèves, Xavier; Cohen-Salmon, Martine

    2015-11-10

    In the brain, most of the vascular system consists of a selective barrier, the blood-brain barrier (BBB) that regulates the exchange of molecules and immune cells between the brain and the blood. Moreover, the huge neuronal metabolic demand requires a moment-to-moment regulation of blood flow. Notably, abnormalities of these regulations are etiological hallmarks of most brain pathologies; including glioblastoma, stroke, edema, epilepsy, degenerative diseases (ex: Parkinson's disease, Alzheimer's disease), brain tumors, as well as inflammatory conditions such as multiple sclerosis, meningitis and sepsis-induced brain dysfunctions. Thus, understanding the signaling events modulating the cerebrovascular physiology is a major challenge. Much insight into the cellular and molecular properties of the various cell types that compose the cerebrovascular system can be gained from primary culture or cell sorting from freshly dissociated brain tissue. However, properties such as cell polarity, morphology and intercellular relationships are not maintained in such preparations. The protocol that we describe here is designed to purify brain vessel fragments, whilst maintaining structural integrity. We show that isolated vessels consist of endothelial cells sealed by tight junctions that are surrounded by a continuous basal lamina. Pericytes, smooth muscle cells as well as the perivascular astrocyte endfeet membranes remain attached to the endothelial layer. Finally, we describe how to perform immunostaining experiments on purified brain vessels.

  1. Collapsible Cryogenic Storage Vessel Project

    Science.gov (United States)

    Fleming, David C.

    2002-01-01

    Collapsible cryogenic storage vessels may be useful for future space exploration missions by providing long-term storage capability using a lightweight system that can be compactly packaged for launch. Previous development efforts have identified an 'inflatable' concept as most promising. In the inflatable tank concept, the cryogen is contained within a flexible pressure wall comprised of a flexible bladder to contain the cryogen and a fabric reinforcement layer for structural strength. A flexible, high-performance insulation jacket surrounds the vessel. The weight of the tank and the cryogen is supported by rigid support structures. This design concept is developed through physical testing of a scaled pressure wall, and through development of tests for a flexible Layered Composite Insulation (LCI) insulation jacket. A demonstration pressure wall is fabricated using Spectra fabric for reinforcement, and burst tested under noncryogenic conditions. An insulation test specimens is prepared to demonstrate the effectiveness of the insulation when subject to folding effects, and to examine the effect of compression of the insulation under compressive loading to simulate the pressure effect in a nonrigid insulation blanket under the action atmospheric pressure, such as would be seen in application on the surface of Mars. Although pressure testing did not meet the design goals, the concept shows promise for the design. The testing program provides direction for future development of the collapsible cryogenic vessel concept.

  2. 50 CFR 660.305 - Vessel identification.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Vessel identification. 660.305 Section 660.305 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND ATMOSPHERIC... Fisheries § 660.305 Vessel identification. (a) Display. The operator of a vessel that is over 25 ft (7.6 m...

  3. 50 CFR 660.704 - Vessel identification.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Vessel identification. 660.704 Section 660.704 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND ATMOSPHERIC... § 660.704 Vessel identification. (a) General. This section only applies to commercial fishing vessels...

  4. 50 CFR 660.504 - Vessel identification.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Vessel identification. 660.504 Section 660.504 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND ATMOSPHERIC... § 660.504 Vessel identification. (a) Official number. Each fishing vessel subject to this subpart must...

  5. 50 CFR 665.16 - Vessel identification.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Vessel identification. 665.16 Section 665... identification. (a) Applicability. Each fishing vessel subject to this part, except those identified in paragraph (e) of this section, must be marked for identification purposes, as follows: (1) A vessel that is...

  6. Development of a New Class of Fe-3Cr-W(V)Ferritic STeels for Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Jawad, M.

    2005-03-31

    The project described in this report dealt with improving the materials performance and fabrication for hydrotreating reactor vessels, heat recovery systems, and other components for the petroleum and chemical industries. The petroleum and chemical industries use reactor vessels that can approach ship weights of approximately 300 tons with vessel wall thicknesses of 3-8 in. These vessels are typically fabricated from Fe-Cr-Mo steels with chromium ranging from 1.25 to 12% and molybdenum from 1 to 2%. Steels in this composition range have great advantages of high thermal conductivity, low thermal expansion, low cost, and good properties obtainable by heat treatment. With all of the advantages of Fe-Cr-Mo steels, several issues are faced in design and fabrication of vessels and related components. These issues include the following: 1. The low strengths of current alloys require thicker sections. 2. Increased thickness causes heat-treatment issues related to nonuniformity across the thickness and thus a failure to achieve optimum properties. 3. Fracture toughness (ductile-to-brittle transition) is a critical safety issue for these vessels, especially in thick sections because of the nonuniformity of the microstructure. 4. The postweld heat treatment (PWHT) needed after welding makes fabrication more timeconsuming with increased cost. 5. PWHT needed after welding also limits any modifications of the large vessels in service. The goal of this project was to reduce the weight of large-pressure-vessel components (ranging from 100 to 300 tons) by approximately 25%, reduce fabrication cost, and improve in-service modification feasibility through development of Fe-3Cr-W(V) steels with a combination of nearly a 50% higher strength, a lower ductile-brittle transition temperature (DBTT), a higher upper-shelf energy, ease of heat treating, and a strong potential for not requiring PWHT.

  7. Microstructure influence on fatigue behaviour of austenitic stainless steels with high molybdenum content; Influencia de la microestructura en el comportamiento a fatiga de aceros inoxidables austeniticos con alto contenido en molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Onoro, J.; Gamboa, R.; Ranninger, C.

    2006-07-01

    Austenitic stainless steels with molybdenum present high mechanical properties and corrosion resistance to aggressive environments. These steels have been used to tank and vessel components for high liquids as phosphoric, nitric and sulphuric acids. These materials with low carbon and nitrogen addition have been proposed candidates as structural materials for the international thermonuclear experimental reactor (ITER) in-vessel components. Molybdenum addition in austenitic stainless steel improves mechanical and corrosion properties, but with it can produce the presence of nitrogen microstructure modifications by presence or precipitation of second phases. This paper summarises the fatigue and corrosion fatigue behaviour of two 317LN stainless steels with different microstructure. Fully austenitic steel microstructure show better fatigue, corrosion fatigue resistance and better ductility than austenitic steel with delta ferrite microstructure, mainly at low stresses. (Author)

  8. Integrating Multiple Autonomous Underwater Vessels, Surface Vessels and Aircraft into Oceanographic Research Vessel Operations

    Science.gov (United States)

    McGillivary, P. A.; Borges de Sousa, J.; Martins, R.; Rajan, K.

    2012-12-01

    Autonomous platforms are increasingly used as components of Integrated Ocean Observing Systems and oceanographic research cruises. Systems deployed can include gliders or propeller-driven autonomous underwater vessels (AUVs), autonomous surface vessels (ASVs), and unmanned aircraft systems (UAS). Prior field campaigns have demonstrated successful communication, sensor data fusion and visualization for studies using gliders and AUVs. However, additional requirements exist for incorporating ASVs and UASs into ship operations. For these systems to be optimally integrated into research vessel data management and operational planning systems involves addressing three key issues: real-time field data availability, platform coordination, and data archiving for later analysis. A fleet of AUVs, ASVs and UAS deployed from a research vessel is best operated as a system integrated with the ship, provided communications among them can be sustained. For this purpose, Disruptive Tolerant Networking (DTN) software protocols for operation in communication-challenged environments help ensure reliable high-bandwidth communications. Additionally, system components need to have considerable onboard autonomy, namely adaptive sampling capabilities using their own onboard sensor data stream analysis. We discuss Oceanographic Decision Support System (ODSS) software currently used for situational awareness and planning onshore, and in the near future event detection and response will be coordinated among multiple vehicles. Results from recent field studies from oceanographic research vessels using AUVs, ASVs and UAS, including the Rapid Environmental Picture (REP-12) cruise, are presented describing methods and results for use of multi-vehicle communication and deliberative control networks, adaptive sampling with single and multiple platforms, issues relating to data management and archiving, and finally challenges that remain in addressing these technological issues. Significantly, the

  9. History of ultrahigh carbon steels

    Energy Technology Data Exchange (ETDEWEB)

    Wadsworth, J.; Sherby, O.D.

    1997-06-20

    The history and development of ultrahigh carbon steels (i.e., steels containing between 1 and 2.l percent C and now known as UHCS) are described. The early use of steel compositions containing carbon contents above the eutectoid level is found in ancient weapons from around the world. For example, both Damascus and Japanese sword steels are hypereutectoid steels. Their manufacture and processing is of interest in understanding the role of carbon content in the development of modern steels. Although sporadic examples of UHCS compositions are found in steels examined in the early part of this century, it was not until the mid-1970s that the modern study began. This study had its origin in the development of superplastic behavior in steels and the recognition that increasing the carbon content was of importance in developing that property. The compositions that were optimal for superplasticity involved the development of steels that contained higher carbon contents than conventional modern steels. It was discovered, however, that the room temperature properties of these compositions were of interest in their own right. Following this discovery, a period of intense work began on understanding their manufacture, processing, and properties for both superplastic forming and room temperature applications. The development of superplastic cast irons and iron carbides, as well as those of laminated composites containing UHCS, was an important part of this history.

  10. Recent Natural Corrosion Inhibitors for Mild Steel: An Overview

    Directory of Open Access Journals (Sweden)

    Marko Chigondo

    2016-01-01

    Full Text Available Traditionally, reduction of corrosion has been managed by various methods including cathodic protection, process control, reduction of the metal impurity content, and application of surface treatment techniques, as well as incorporation of suitable alloys. However, the use of corrosion inhibitors has proven to be the easiest and cheapest method for corrosion protection and prevention in acidic media. These inhibitors slow down the corrosion rate and thus prevent monetary losses due to metallic corrosion on industrial vessels, equipment, or surfaces. Inorganic and organic inhibitors are toxic and costly and thus recent focus has been turned to develop environmentally benign methods for corrosion retardation. Many researchers have recently focused on corrosion prevention methods using green inhibitors for mild steel in acidic solutions to mimic industrial processes. This paper provides an overview of types of corrosion, corrosion process, and mainly recent work done on the application of natural plant extracts as corrosion inhibitors for mild steel.

  11. Modelling Steel Behaviour

    OpenAIRE

    Anderberg, Yngve

    1986-01-01

    When modelling material mechanical behaviour, an analytical description is required of the relationship between stresses and strains. A computer oriented mechanical behaviour model for steel is described. The model is based on the fact that the deformation process at transient high temperature conditions can be desribed by three strain components which are separately found in different steady state tests. It is shown that a behaviour model based on steady state data satisfactorily predicts be...

  12. Progress of ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

    2013-10-15

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

  13. Wootz Damascus steel blades

    Energy Technology Data Exchange (ETDEWEB)

    Verhoeven, J.D.; Gibson, E.D. [Ames Lab., IA (United States); Pendray, A.H. [ABS Master Bladesmith, Williston, FL (United States)

    1996-07-01

    Wootz Damascus steel blades contain surface patterns produced by bands of cementite particles which are generated in situ as the blades are forged from small ingots. A process for making these blades has recently been developed which involves making ingots in a gas-fired furnace followed by forging to blade shapes. This study presents a series of additional experiments which provide strong evidence that the mechanism responsible for the formation of the aligned cementite bands is similar to the mechanism that produces banded hypoeutectoid steels. That mechanism attributes the selective formation of ferrite bands to microsegregated alloying elements. The results of this study show that the cementite bands will form in ultraclean hypereutectoid steels (P and S levels <0.003 wt. %) by the addition of small amounts of carbide-forming elements V, Cr, and Ti at a combined level of <0.02 wt. %. The results present strong evidence that the cementite bands are formed by a selective coarsening of cementite particles during the thermal cycling of the forging process. The particle coarsening is induced to occur preferentially in the interdendritic regions of the alloys by the very small additions of the carbide-forming elements.

  14. Fracture Mechanisms in Steel Castings

    Directory of Open Access Journals (Sweden)

    Stradomski Z.

    2013-09-01

    Full Text Available The investigations were inspired with the problem of cracking of steel castings during the production process. A single mechanism of decohesion - the intergranular one - occurs in the case of hot cracking, while a variety of structural factors is decisive for hot cracking initiation, depending on chemical composition of the cast steel. The low-carbon and low-alloyed steel castings crack due to the presence of the type II sulphides, the cause of cracking of the high-carbon tool cast steels is the net of secondary cementite and/or ledeburite precipitated along the boundaries of solidified grains. Also the brittle phosphor and carbide eutectics precipitated in the final stage solidification are responsible for cracking of castings made of Hadfield steel. The examination of mechanical properties at 1050°C revealed low or very low strength of high-carbon cast steels.

  15. The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-01

    MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of

  16. Synthesis of thermit noncorrodible steels

    OpenAIRE

    Жигуц, Юрій Юрійович

    2013-01-01

    The present paper the basic solutions to the problem of obtaining cavitation-resistant steels examined the use of thermite steels, the benefits of combining thermite steels with metallotermic methods of getting is showed. The advantages of metallotermic synthesis methods include: autonomy of processes, independence of energy sources, simplicity of equipment, high-performance process and easy transition from experimental research to industrial production. The need to developed the technology o...

  17. Thermal energy storage using Prestressed Cast Iron Vessels (PCIV). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Gilli, P.V.; Beckmann, G.; Schilling, F.E.

    1977-06-01

    The wide-spread application of thermal energy and high-pressure air storage to electric power generation has so far been hampered by the lack of large high-pressure storage vessels of reasonable cost. Welded steel vessels are too expensive for this purpose. However, the Prestressed Cast Iron Vessel (PCIV), developed as a nuclear reactor pressure vessel by Siempelkamp Giesserei KG of Krefeld, FRG, has the potential of complying with these requirements. Applications of the PCIV include: high-pressure air storage for the quick start-up of open cycle gas turbines; pressurized high-temperature sensible heat storage by means of solids with a gaseous heat transfer medium for closed cycle gas turbines of future solar power stations; and pressurized hot water storage for nuclear, solar, or coal-fired steam power plants, employing either separate peaking turbines or overloadable main turbine sets. A reference PCIV of 8000 m/sup 3/, 275/sup 0/C, with hot going walls and cold going tendons was developed, designed, and stress-analysed. A parametric study showed that pressures between 4 and 8 MPa and L/D ratios larger than 4 should be optimal. Cost of the reference vessel is about $10,000,000 or 33 to 50 $/kWh electric energy stored. Cost of peak power will be from 30 to 100 mills/kWh, depending on many parameters.

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

    Energy Technology Data Exchange (ETDEWEB)

    Habil; Ziliukas, A.

    1997-04-01

    In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

  19. 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.

  20. Cold weld cracking susceptibility of high strength low alloyed (HSLA steel NIONIKRAL 70

    Directory of Open Access Journals (Sweden)

    A. S. Tawengi

    2014-10-01

    Full Text Available In view of the importance of high strength low alloy (HSLA steels, particularly for critical applications such as offshore plat forms, pipeline and pressure vessels, this paper reports on an investigation of how to weld this type of steel without cold cracking. Using manual metal arc welding process and Tekken test (Y - Grove test has been carried out both to observe the cold cracking phenome non, and to investigate the influencing factors, such as preheating temperature and energy input, as well as electrode strength and diameter. How ever the results of the experiments show that there is a risk of cold cracking.

  1. Exponential Stabilization of an Underactuated Surface Vessel

    Directory of Open Access Journals (Sweden)

    Kristin Y. Pettersen

    1997-07-01

    Full Text Available The paper shows that a large class of underactuated vehicles cannot be asymptotically stabilized by either continuous or discontinuous state feedback. Furthermore, stabilization of an underactuated surface vessel is considered. Controllability properties of the surface vessels is presented, and a continuous periodic time-varying feedback law is proposed. It is shown that this feedback law exponentially stabilizes the surface vessel to the origin, and this is illustrated by simulations.

  2. Steel designers' handbook

    CERN Document Server

    Gorenc, Branko; Tinyou, Ron

    2012-01-01

    The Revised 7th Edition of Steel Designers' Handbook is an invaluable tool for all practising structural, civil and mechanical engineers as well as engineering students at university and TAFE in Australia and New Zealand. It has been prepared in response to changes in the design Standard AS 4100, the structural Design Actions Standards, AS /ANZ 1170, other processing Standards such as welding and coatings, updated research as well as feedback from users. This edition is based on Australian Standard (AS) 4100: 1998 and subsequent amendments. The worked numerical examples in the book have been e

  3. Typhoon of Steel

    OpenAIRE

    Hamamoto, Gena

    2012-01-01

    Typhoon of Steel is a short community-based documentary film that explores the lives of two Okinawan American Kibei Nisei who served in the U.S. military as linguists in the Battle of Okinawa during World War II. While Japanese Americans on the West Coast were incarcerated in camps, these men risked their lives to prove their loyalty to America. Born in the U.S. and raised in Okinawa, their cultural and linguistic skills were a tactical asset to the military. But emotions ran high as they ...

  4. Mechanics in Steels through Microscopy

    NARCIS (Netherlands)

    Zandbergen, H.W.; Tirumalasetty, G.K.

    The goal of the study consolidated in this thesis is to understand the mechanics in steels using microscopy. In particular, the mechanical response of Transformation Induced Plasticity (TRIP) steels is correlated with their microstructures. Chapter 1 introduces the current state of the art of TRIP

  5. corrosion inhibitor for carbon steels

    African Journals Online (AJOL)

    potentiodynamic polarisation techniques. It was found that. CNSL reduces the extent of the electrochemical processes taking place on carbon steel undergoing corrosion. The corrosion rate of the carbon steel was reduced by over 92 % when only 300 ppm of CNSL was applied. This indicates that. CNSL is a potential ...

  6. Mechanics in Steels through Microscopy

    NARCIS (Netherlands)

    Tirumalasetty, G.K.

    2013-01-01

    The goal of the study consolidated in this thesis is to understand the mechanics in steels using microscopy. In particular, the mechanical response of Transformation Induced Plasticity (TRIP) steels is correlated with their microstructures. Chapter 1 introduces the current state of the art of TRIP

  7. Heavy-Section Steel Technology Program quarterly progress report, January-March 1980

    Energy Technology Data Exchange (ETDEWEB)

    Whitman, G.D.; Bryan, R.H.

    1980-07-01

    The program comprises studies related to all areas of the technology of materials fabricated into thick-section primary-coolant containment systems of light-water-cooled nuclear power reactors. The principal area of investigation is the behavior and structural integrity of steel pressure vessels containing cracklike flaws. Current work is organized into the following tasks: (1) program administration and procurement, (2) fracture mechanics analyses and investigations, (3) investigations of irradiated materials, (4) thermal shock investigations, and (5) pressure vessel investigations. Work performed under the existing research and development subcontracts is included in this report.

  8. Investigation of the deformation of in-vessel components of a nuclear fusion experiment using optical strain sensors

    Energy Technology Data Exchange (ETDEWEB)

    Vorpahl, Christian Georg

    2013-05-03

    A fibre-optic, EM-insensitive measurement for the deformation of in-vessel components has successfully been installed and operated at the nuclear fusion experiment ASDEX Upgrade. The sensors were tested for their neutron tolerance and vacuum compatibility. Installation was done by copper-steel laser beam welding. Measurements of in-service oscillations due to all three existing types of load cases show good agreement with theory and simulations. A fatigue lifetime assessment was performed.

  9. Dismantling of the reactor pressure vessel internals in the NPP Wuergassen; Zerlegung der RDG- Einbauten im KKW Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Bruhn, Jan Hendrik [AREVA NP GmbH (Germany)

    2008-07-01

    The reactor pressure vessel internals of the NPP Wuergassen were dismantled and dissected in the time period 19-03-2007 to 21-09-2007 and packaged for final disposal. The total amount was about 18 tons of steel. The dissected was performed using specific water-quartz sand-cutting tools within a settling tank. The dismantling concept intended to cut shape-optimized pieces with respect to space saving packaging. The project included a specific water treatment system.

  10. Steels from materials science to structural engineering

    CERN Document Server

    Sha, Wei

    2013-01-01

    Steels and computer-based modelling are fast growing fields in materials science as well as structural engineering, demonstrated by the large amount of recent literature. Steels: From Materials Science to Structural Engineering combines steels research and model development, including the application of modelling techniques in steels.  The latest research includes structural engineering modelling, and novel, prototype alloy steels such as heat-resistant steel, nitride-strengthened ferritic/martensitic steel and low nickel maraging steel.  Researchers studying steels will find the topics vital to their work.  Materials experts will be able to learn about steels used in structural engineering as well as modelling and apply this increasingly important technique in their steel materials research and development. 

  11. New sacrificial material for ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Komlev, Andrei A., E-mail: komlev@kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Nuclear Power Safety Division, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Almjashev, Vyacheslav I., E-mail: vac@mail.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Bechta, Sevostian V., E-mail: bechta@safety.sci.kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Khabensky, Vladimir B., E-mail: vladimirkhabensky@gmail.com [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Granovsky, Vladimir S., E-mail: gran@niti.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Gusarov, Victor V., E-mail: victor.v.gusarov@gmail.com [Ioffe Institute, 26 Polytekhnicheskaya Str., St. Petersburg, 194021 (Russian Federation)

    2015-12-15

    A new functional (sacrificial) material has been developed in the Fe{sub 2}O{sub 3}–SrO–Al{sub 2}O{sub 3}–CaO system based on strontium hexaferrite ceramic in concrete matrix. The method of producing SM has been advanced technologically; this technological effectiveness allows the SM to be used in ex-vessel core catchers with corium spreading as well as in crucible-type core catchers. Critical properties regarding the efficiency of SM in ex-vessel core catchers, such as porosity, pycnometric density, apparent density, solidus and liquidus temperatures, and water content have been measured. Suitable fractions of SrFe{sub 12}O{sub 19} and high alumina cement (HAC) were found in the SM based on thermodynamic analysis of the SM/corium interaction. The use of sacrificial steel as an additional heat adsorption component in the core catcher allowed us to increase the mass fraction range of SrFe{sub 12}O{sub 19} in the SM from 0.3−0.5 to 0.3–0.85. The activation temperature of the SM/corium interaction has been shown to correspond to the liquidus temperature of the local composition at the SM/corium interface. The calculated value of this temperature was 1716 °C. Analysis of phase transformations in the SrO–Fe{sub 2}O{sub 3} system revealed advantages of the SrFe{sub 12}O{sub 19}–based sacrificial material compared with the Fe{sub 2}O{sub 3}-contained material owing to the time proximity of SrFe{sub 12}O{sub 19} decomposition and corium interaction activation. - Highlights: • A sacrificial material (SM) was developed for ex-vessel core catcher. • Suitable proportions in the SrFe{sub 12}O{sub 19}–Al{sub 2}O{sub 3}·CaO–Fe system were determined. • Hydrogen release limitation was shown for ex-vessel corium retention with the SM. • Calculated temperature of the active initiation of corium/SM interaction is 1716 °C. • Functional properties of the SM were measured.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-15

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

  14. Vessel classification method based on vessel behavior in the port of Rotterdam

    NARCIS (Netherlands)

    Zhou, Y.; Daamen, W.; Vellinga, T.; Hoogendoorn, S.P.

    2015-01-01

    AIS (Automatic Identification System) data have proven to be a valuable source to investigate vessel behavior. The analysis of AIS data provides a possibility to recognize vessel behavior patterns in a waterway area. Furthermore, AIS data can be used to classify vessel behavior into several

  15. A computational algorithm addressing how vessel length might depend on vessel diameter

    Science.gov (United States)

    Jing Cai; Shuoxin Zhang; Melvin T. Tyree

    2010-01-01

    The objective of this method paper was to examine a computational algorithm that may reveal how vessel length might depend on vessel diameter within any given stem or species. The computational method requires the assumption that vessels remain approximately constant in diameter over their entire length. When this method is applied to three species or hybrids in the...

  16. 76 FR 59660 - Proposed Information Collection; Comment Request; Permitting, Vessel Identification, and Vessel...

    Science.gov (United States)

    2011-09-27

    ...; Permitting, Vessel Identification, and Vessel Monitoring System Requirements for the Commercial Bottomfish... compliance with federal identification requirements and carry and maintain a satellite- based vessel monitoring system (VMS). This collection of information is needed for permit issuance, to identify actual or...

  17. Analysis of the structural steels corrosion resistance in sour water from petroleum refineries; Analise da resistencia a corrosao de acos estruturais em aguas acidas de refinarias de petroleo

    Energy Technology Data Exchange (ETDEWEB)

    Proenca, Marcos B.; Freire, Celia M. de A. [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica; Santos, Margatita B. [Universidade Estadual de Campinas, SP (Brazil). Inst. de Fisica. Dept. de Fisica Aplicada

    1994-07-01

    The presence of H{sub 2} S in refineries FCC sour water provokes the formation of a passive Fe S scale. The cyanides present on sour water remove this scale, raising the corrosion rate in pipping and vessels. In this work it was measured the corrosion rate of structural steels in this water by electrochemical methods. Anodic polarization curves were plotted and the corrosion rates of the steels were determined. (author)

  18. Protective coating of austenitic steel using robotized GMAW temper-bead technique; Rechargement d'inox austenitique en MAG temperbead robotise

    Energy Technology Data Exchange (ETDEWEB)

    Carpreau, J.M. [Electricite de France (EDF/R and D), Recherche et Developpement, 92 - Chatou (France); Dainelli, P. [Institut de Soudure, 57 - Yutz (France)

    2009-07-15

    This paper summarises experimental results obtained in a study of GMAW temper-bead on low alloyed steel with austenitic consumables. Temper-bead on low alloyed steel with austenitic consumables is mainly used for repairing operations of heavy components such as vessel reactor of nuclear power plants. Experimental work aims at showing the performance of GMAW compared to GTAW and the consequences of GMAW temper-bead on 2OMND5 heat affected zones. (authors)

  19. 15 CFR 970.205 - Vessel safety.

    Science.gov (United States)

    2010-01-01

    ... following: (1) That any foreign flag vessel whose flag state is party to the International Convention for Safety of Life at Sea, 1974 (SOLAS 74) possesses current valid SOLAS 74 certificates; (2) That any foreign flag vessel whose flag state is not party to SOLAS 74 but is party to the International Convention...

  20. Assessing Vessel Traffic Service Operator Situation Awareness

    NARCIS (Netherlands)

    Wiersma, J.W.F.

    2010-01-01

    This thesis describes my study of situation awareness assessment of Vessel Traffic Service (VTS) operators. VTS operators are the traffic controllers on the water. They are responsible for a safe and efficient handling of vessel traffic. They monitor traffic, provide information on request and

  1. Hereditary cerebral small vessel disease and stroke

    DEFF Research Database (Denmark)

    Søndergaard, Christian Baastrup; Nielsen, Jørgen Erik; Hansen, Christine Krarup

    2017-01-01

    Cerebral small vessel disease is considered hereditary in about 5% of patients and is characterized by lacunar infarcts and white matter hyperintensities on MRI. Several monogenic hereditary diseases causing cerebral small vessel disease and stroke have been identified. The purpose of this system...

  2. 33 CFR 401.67 - Explosive vessels.

    Science.gov (United States)

    2010-07-01

    ... TRANSPORTATION SEAWAY REGULATIONS AND RULES Regulations Dangerous Cargo § 401.67 Explosive vessels. A vessel carrying explosives, either Government or commercial, as defined in the Dangerous Cargo Act of the United States and in the International Maritime Dangerous Goods Code, Class 1, Divisions 1.1 to 1.5 inclusive...

  3. Analyzing Vessel Behavior Using Process Mining

    NARCIS (Netherlands)

    Maggi, F.M.; Mooij, A.J.; Aalst, W.M.P. van der

    2013-01-01

    In the maritime domain, electronic sensors such as AIS receivers and radars collect large amounts of data about the vessels in a certain geographical area. We investigate the use of process mining techniques for analyzing the behavior of the vessels based on these data. In the context of maritime

  4. 78 FR 39649 - Passenger Vessels Accessibility Guidelines

    Science.gov (United States)

    2013-07-02

    ... TRANSPORTATION BARRIERS COMPLIANCE BOARD 36 CFR Part 1196 RIN 3014-AA11 Passenger Vessels Accessibility... Tuesday, June 25, 2013, make the following correction: PART 1196--PASSENGER VESSELS ACCESSIBILITY... ``Figure V703.7.2.1 International Symbol of Accessibility'' and are added to read as set forth below...

  5. Review of environmental effects on fatigue crack growth of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Shack, W.J.; Kassner, T.F. [Argonne National Lab., IL (United States)

    1994-05-01

    Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry.

  6. Method for the Calculation of DPA in the Reactor Pressure Vessel of Atucha II

    Directory of Open Access Journals (Sweden)

    J. A. Mascitti

    2011-01-01

    It was determined that the maximum DPA rate in the RPV wall with fresh fuel element (FE is 3.76(3 × 10-12 s-1, it takes place in front of FEs BA42 and BL43, and it is symmetrical about the central channel, LG04, and LH03.

  7. Nuclear reactor vessel fuel thermal insulating barrier

    Science.gov (United States)

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

    2013-03-19

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

  8. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  9. 49 CFR 192.55 - Steel pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Steel pipe. 192.55 Section 192.55 Transportation... BY PIPELINE: MINIMUM FEDERAL SAFETY STANDARDS Materials § 192.55 Steel pipe. (a) New steel pipe is... in accordance with paragraph (c) or (d) of this section. (b) Used steel pipe is qualified for use...

  10. Improving the toughness of ultrahigh strength steel

    Energy Technology Data Exchange (ETDEWEB)

    Soto, Koji [Univ. of California, Berkeley, CA (United States)

    2002-01-01

    The ideal structural steel combines high strength with high fracture toughness. This dissertation discusses the toughening mechanism of the Fe/Co/Ni/Cr/Mo/C steel, AerMet 100, which has the highest toughness/strength combination among all commercial ultrahigh strength steels. The possibility of improving the toughness of this steel was examined by considering several relevant factors.

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

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  12. Methods of making bainitic steel materials

    Energy Technology Data Exchange (ETDEWEB)

    Bakas, Michael Paul; Chu, Henry Shiu-Hung; Zagula, Thomas Andrew; Langhorst, Benjamin Robert

    2018-01-16

    Methods of making bainitic steels may involve austenitizing a quantity of steel by exposing the quantity of steel to a first temperature. A composition of the quantity of steel may be configured to impede formation of non-bainite ferrite, pearlite, and Widmanstatten ferrite. The quantity of steel may be heat-treated to form bainite by exposing the quantity of steel to a second, lower temperature. The second, lower temperature may be stabilized by exposing the quantity of steel to the second, lower temperature in the presence of a thermal ballast.

  13. Stahlschüssel key to steel

    CERN Document Server

    Wegst, W S

    2016-01-01

    The Key to Steel (Stahlschlüssel/Stahlschluessel) cross reference book will help you to decode / decipher steel designations and find equivalent materials worldwide. The 2016 edition includes more than 70,000 standard designations and trade names from approximately 300 steelmakers and suppliers. Presentation is trilingual: English, French, and German. Materials covered include structural steels, tool steels, valve steels, high temperature steels and alloys, stainless and heat-resisting steels, and more. Standards and designations from 25 countries are cross-referenced.

  14. 46 CFR 535.312 - Vessel charter party-exemption.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 9 2010-10-01 2010-10-01 false Vessel charter party-exemption. 535.312 Section 535.312... Vessel charter party-exemption. (a) For purposes of this section, vessel charter party shall mean a... operational limitations, if any) under which the vessel will be employed. (b) Vessel charter parties, as...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-07-01

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

  16. Modelling fracture in ferritic steel

    CERN Document Server

    Smith, G

    2002-01-01

    Results from mathematical models and computer simulations of fracture in polycrystalline steels are presented for a range of temperatures. The proportions of intergranular and intragranular failure predicted are compared with experimental results for brittle fracture, ductile fracture and in the transition region. Interactive software to create two-dimensional polycrystalline models, which allow a range of physical to be varied independently, is described. The results include those for model materials chosen to match steels used by the power generation industry. The models simulate segregation and cavitation effects in steel and fracture of weldments and their associated heat-affected zones.

  17. Cold-formed steel design

    CERN Document Server

    Yu, Wei-Wen

    2010-01-01

    The definitive text in the field, thoroughly updated and expanded Hailed by professionals around the world as the definitive text on the subject, Cold-Formed Steel Design is an indispensable resource for all who design for and work with cold-formed steel. No other book provides such exhaustive coverage of both the theory and practice of cold-formed steel construction. Updated and expanded to reflect all the important developments that have occurred in the field over the past decade, this Fourth Edition of the classic text provides you with more of the detailed, up-to-the-minute techni

  18. Steel fiber replacement of mild steel in prestressed concrete beams

    Science.gov (United States)

    2010-10-01

    In traditional prestressed concrete beams, longitudinal prestressed tendons serve to resist bending moment and : transverse mild steel bars (or stirrups) are used to carry shear forces. However, traditional prestressed concrete I-beams : exhibit earl...

  19. High-strength, low-alloy steels.

    Science.gov (United States)

    Rashid, M S

    1980-05-23

    High-strength, low-alloy (HSLA) steels have nearly the same composition as plain carbon steels. However, they are up to twice as strong and their greater load-bearing capacity allows engineering use in lighter sections. Their high strength is derived from a combination of grain refinement; precipitation strengthening due to minor additions of vanadium, niobium, or titanium; and modifications of manufacturing processes, such as controlled rolling and controlled cooling of otherwise essentially plain carbon steel. HSLA steels are less formable than lower strength steels, but dualphase steels, which evolved from HSLA steels, have ferrite-martensite microstructures and better formability than HSLA steels of similar strength. This improved formability has substantially increased the utilization potential of high-strength steels in the manufacture of complex components. This article reviews the development of HSLA and dual-phase steels and discusses the effects of variations in microstructure and chemistry on their mechanical properties.

  20. 2169 steel waveform experiments.

    Energy Technology Data Exchange (ETDEWEB)

    Furnish, Michael David; Alexander, C. Scott; Reinhart, William Dodd; Brown, Justin L.

    2012-11-01

    In support of LLNL efforts to develop multiscale models of a variety of materials, we have performed a set of eight gas gun impact experiments on 2169 steel (21% Cr, 6% Ni, 9% Mn, balance predominantly Fe). These experiments provided carefully controlled shock, reshock and release velocimetry data, with initial shock stresses ranging from 10 to 50 GPa (particle velocities from 0.25 to 1.05 km/s). Both windowed and free-surface measurements were included in this experiment set to increase the utility of the data set, as were samples ranging in thickness from 1 to 5 mm. Target physical phenomena included the elastic/plastic transition (Hugoniot elastic limit), the Hugoniot, any phase transition phenomena, and the release path (windowed and free-surface). The Hugoniot was found to be nearly linear, with no indications of the Fe phase transition. Releases were non-hysteretic, and relatively consistent between 3- and 5-mmthick samples (the 3 mm samples giving slightly lower wavespeeds on release). Reshock tests with explosively welded impactors produced clean results; those with glue bonds showed transient releases prior to the arrival of the reshock, reducing their usefulness for deriving strength information. The free-surface samples, which were steps on a single piece of steel, showed lower wavespeeds for thin (1 mm) samples than for thicker (2 or 4 mm) samples. A configuration used for the last three shots allows release information to be determined from these free surface samples. The sample strength appears to increase with stress from ~1 GPa to ~ 3 GPa over this range, consistent with other recent work but about 40% above the Steinberg model.