Metallurgy of steels for PWR pressure vessels

International Nuclear Information System (INIS)

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

1980-01-01

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

Metallurgy of steels for PWR pressure vessels

Energy Technology Data Exchange (ETDEWEB)

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

1980-09-01

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

Development of High Heat Input Welding High Strength Steel Plate for Oil Storage Tank in Xinyu Steel Company

Science.gov (United States)

Zhao, Hemin; Dong, Fujun; Liu, Xiaolin; Xiong, Xiong

This essay introduces the developed high-heat input welding quenched and tempered pressure vessel steel 12MnNiVR for oil storage tank by Xinyu Steel, which passed the review by the Boiler and Pressure Vessel Standards Technical Committee in 2009. The review comments that compared to the domestic and foreign similar steel standard, the key technical index of enterprise standard were in advanced level. After the heat input of 100kJ/cm electro-gas welding, welded points were still with excellent low temperature toughness at -20°C. The steel plate may be constructed for oil storage tank, which has been permitted by thickness range from 10 to 40mm, and design temperature among -20°C-100°C. It studied microstructure genetic effects mechanical properties of the steel. Many production practices indicated that the mechanical properties of products and the steel by stress relief heat treatment of steel were excellent, with pretreatment of hot metal, converter refining, external refining, protective casting, TMCP and heat treatment process measurements. The stability of performance and matured technology of Xinyu Steel support the products could completely service the demand of steel constructed for 10-15 million cubic meters large oil storage tank.

Dynamic fracture characterization of a pressure vessel steel

International Nuclear Information System (INIS)

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

1991-01-01

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

A crack arrest test using a toughness gradient steel plate

International Nuclear Information System (INIS)

Okamura, H.; Yagawa, G.; Urabe, Y.; Satoh, M.; Sano, J.

1995-01-01

Pressurized thermal shock (PTS) is a phenomenon that can occur in the reactor pressure vessels (RPVs) with internal pressure and is one of the most severe stress conditions that can be applied to the vessel. Preliminary research has shown that no PTS concern is likely to exist on Japanese RPVs during their design service lives. However, public acceptance of vessel integrity requires analyses and experiment in order to establish an analytical method and a database for life extension of Japanese RPVs. The Japanese PTS integrity study was carried out from FY 1983 to FY 1991 as a national project by Japan Power Engineering and Inspection Corporation (JAPEIC) under contract with Ministry of International Trade and Industry (MITI) in cooperation with LWR utilities and vendors. Here, a crack arrest test was carried out using a toughness gradient steel plate with three layers to study the concept of crack arrest toughness. Four-point bending load with thermal shock was applied to the large flat plate specimen with a surface crack. Five crack initiations and arrests were observed during the test and the propagated crack bifurcated. Finally, cracks were arrested at the boundary of the first and the second layer, except for a small segment of the crack. The first crack initiation took place slightly higher than the lower bound of K Ic data obtained by ITCT specimens. That is, the K IC concept for brittle crack initiation was verified for heavy section steel plates. The first crack arrest took place within the scatter band of K Ia and K Id data for the first layer. That is, the K Ia concept appears applicable for crack arrest of a short crack jump

The reactor vessel steels

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Diffusion zinc plating of structural steels

International Nuclear Information System (INIS)

Kazakovskaya, Tatiana; Goncharov, Ivan; Tukmakov, Victor; Shapovalov, Vyacheslav

2004-01-01

The report deals with the research on diffusion zinc plating of structural steels when replacing their cyanide cadmium plating. The results of the experiments in the open air, in vacuum, in the inert atmosphere, under various temperatures (300 - 500 deg.C) for different steel brands are presented. It is shown that diffusion zinc plating in argon or nitrogen atmosphere ensures obtaining the qualitative anticorrosion coating with insignificant change of mechanical properties of steels. The process is simple, reliable, ecology pure and cost-effective. (authors)

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Comparison of BR3 Surveillance and Vessel Plates to the Surrogate Plates Representative of the Yankee Rowe PWR Vessel

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

The sister pressure vessels at the BR3 and Yankee Rowe PWR plants were operated at lower-than-usual temperature ( 260 degrees Celsius) and their plates were austenitized a higher-than-usual temperature (970 degrees Celsius) - a heat treatment leading to a coarser microstructure than is typical for the fine grain plates considered in development of USNRC Regulatory Guide 1.99. The surveillance programs provided by Westinghouse for the two plants were limited to the same A302-B plate representative of the Rowe vessel upper shell plate; this material displayed outlier behaviour characterized by a 41 J Charpy-V Notch shift significantly larger than predicted by Regulatory Guide 1.99. Because lower irradiation temperature and nickel alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rate plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements embodied in the PTS screening criterion. This paper compares free 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, and was trepanned in early 1995; (3) The accelerated irradiations in the Belgian test reactor BR2 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 contended that the PTS screening criterion was never attained by the BR3 and Rowe plates, and

Comparison of BR3 Surveillance and Vessel Plates to the Surrogate Plates Representative of the Yankee Rowe PWR Vessel

International Nuclear Information System (INIS)

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

1998-07-01

The sister pressure vessels at the BR3 and Yankee Rowe PWR plants were operated at lower-than-usual temperature ( 260 degrees Celsius) and their plates were austenitized a higher-than-usual temperature (970 degrees Celsius) - a heat treatment leading to a coarser microstructure than is typical for the fine grain plates considered in development of USNRC Regulatory Guide 1.99. The surveillance programs provided by Westinghouse for the two plants were limited to the same A302-B plate representative of the Rowe vessel upper shell plate; this material displayed outlier behaviour characterized by a 41 J Charpy-V Notch shift significantly larger than predicted by Regulatory Guide 1.99. Because lower irradiation temperature and nickel alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rate plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements embodied in the PTS screening criterion. This paper compares free 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, and was trepanned in early 1995; (3) The accelerated irradiations in the Belgian test reactor BR2 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 contended that the PTS screening criterion was never attained by the BR3 and Rowe plates, and

Comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe PWR vessel

International Nuclear Information System (INIS)

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

1999-01-01

The sister pressure vessels at the BR3 and Yankee Rowe PWR plants were operated at lower-than-usual temperature (∼260 C) and their plates were austenitized at higher-than-usual temperature (∼970 C) -- a heat treatment leading to a coarser microstructure than is typical for the fine grain plates considered in development of USNRC Regulatory Guide 1.99. The surveillance programs provided by Westinghouse for the two plants were limited to the same A302-B plate representative of the Rowe vessel upper shell plate; this material displayed outlier behavior characterized by a 41J. Charpy-V Notch shift significantly larger than predicted by Regulatory Guide 1.99. Because lower irradiation temperature and nickel 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 embodied 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, and was trepanned in early 1995; (3) The accelerated irradiations in the Belgian test reactor BR2 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 contended that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel

Residual stresses in weld-clad reactor pressure vessel steel

International Nuclear Information System (INIS)

Bertram, W.

1975-01-01

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

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

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

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

International Nuclear Information System (INIS)

Bangash, Y.

1987-01-01

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

East/west steels for reactor pressure vessels

International Nuclear Information System (INIS)

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

1997-01-01

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

Plating on stainless steel alloys

International Nuclear Information System (INIS)

Dini, J.W.; Johnson, H.R.

1981-01-01

Quantitative adhesion data are presented for a variety of electroplated stainless steel type alloys. Results show that excellent adhesion can be obtained by using a Wood's nickel strike or a sulfamate nickel strike prior to final plating. Specimens plated after Wood's nickel striking failed in the deposit rather than at the interface between the substrate and the coating. Flyer plate quantitative tests showed that use of anodic treatment in sulfuric acid prior to Wood's nickel striking even further improved adhesion. In contrast activation of stainless steels by immersion or cathodic treatment in hydrochloric acid resulted in very reduced bond strengths with failure always occurring at the interface between the coating and substrate

Plastic collapse load of corroded steel plates

Indian Academy of Sciences (India)

Keywords. Corroded steel plate; plastic collapse; FEM; rough surface. ... The main aim of present work is to study plastic collapse load of corroded steel plates with irregular surfaces under tension. Non-linear finite element method ... Department of Ocean Engineering, AmirKabir University of Technology, 15914 Tehran, Iran ...

Electroless Plated Nanodiamond Coating for Stainless Steel Passivation

International Nuclear Information System (INIS)

Li, D.; Korinko, P.; Spencer, W.; Stein, E.

2016-01-01

Tritium gas sample bottles and manifold components require passivation surface treatments to minimize the interaction of the hydrogen isotopes with surface contamination on the stainless steel containment materials. This document summarizes the effort to evaluate electroless plated nanodiamond coatings as a passivation layer for stainless steel. In this work, we developed an electroless nanodiamond (ND)-copper (Cu) coating process to deposit ND on stainless steel parts with the diamond loadings of 0%, 25% and 50% v/v in a Cu matrix. The coated Conflat Flanged Vessel Assemblies (CFVAs) were evaluated on surface morphology, composition, ND distribution, residual hydrogen release, and surface reactivity with deuterium. For as-received Cu and ND-Cu coated CFVAs, hydrogen off-gassing is rapid, and the off-gas rates of H 2 was one to two orders of magnitude higher than that for both untreated and electropolished stainless steel CFVAs, and hydrogen and deuterium reacted to form HD as well. These results indicated that residual H 2 was entrapped in the Cu and ND-Cu coated CFVAs during the coating process, and moisture was adsorbed on the surface, and ND and/or Cu might facilitate catalytic isotope exchange reaction for HD formation. However, hydrocarbons (i.e., CH 3 ) did not form, and did not appear to be an issue for the Cu and ND-Cu coated CFVAs. After vacuum heating, residual H 2 and adsorbed H 2 O in the Cu and ND-Cu coated CFVAs were dramatically reduced. The H 2 off-gassing rate after the vacuum treatment of Cu and 50% ND-Cu coated CFVAs was on the level of 10 -14 l mbar/s cm 2 , while H 2 O off-gas rate was on the level of 10 -15 l mbar/s cm 2 , consistent with the untreated or electropolished stainless steel CFVA, but the HD formation remained. The Restek EP bottle was used as a reference for this work. The Restek Electro-Polished (EP) bottle and their SilTek coated bottles tested under a different research project exhibited very little hydrogen off-gassing and

Recent trend of titanium-clad steel plate/sheet (NKK)

International Nuclear Information System (INIS)

Kimura, Hideto

1997-01-01

The roll-bonding process for titanium-clad steel production enabled the on-line manufacturing and quality control of the products which are usually applied for the production of steel plate and sheet by the steel producers. The recent trend of roll-bonded titanium-clad steel which has an excellent corrosion resistance together with the advantage in cost-saving are mainly described in this article as to the demand, production technique and new application aspects. Though the predominant usage of titanium-clad steel plate has been in power-generating plants, enlargeing utilization in the chemical plants such as terephthalic acid production plants is leading the growth in the market of titanium-clad steel plate. Also, the application of titanium-clad steel plates and sheets for the lining the marine structures is expected as one of the best solution to long-term surface protection for their outstanding corrosion resistance against sea water. (author)

The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

International Nuclear Information System (INIS)

Peacock, A.; Girlinger, A.; Vorkoeper, A.; Boscary, J.; Greuner, H.; Hurd, F.; Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G.

2011-01-01

320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m 2 , are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m 2 and a maximum heat load of 200 kW/m 2 with a water velocity of approximately 2 m s -1 . High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m 2 for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

Steel plate reinforcement of orthotropic bridge decks

NARCIS (Netherlands)

Teixeira de Freitas, S.

2012-01-01

The PhD research is focused on the reinforcement of fatigue cracked orthotropic steel bridge decks (OBD) by adding a second steel plate to the existing deck. The main idea is to stiffen the existing deck plate, which will reduce the stresses at the fatigue sensitive details and extend the fatigue

Special heavy plates and steel solutions for bridge building

Science.gov (United States)

Lehnert, Tobias

2017-09-01

In many European countries infrastructure, -road as well as railway infrastructure-, needs intensive investments to follow the growing demands of mobility and goods traffic. Steel or steel composite bridges offer in this context viable and very sustainable solutions. Due to its unlimited recyclability steel can in general be seen as the ideal material for such sustainable constructions, but especially when designers or fabricators exploit the nowadays available possibilities of steel industry very cost-efficient and remarkable constructions are realizable. This paper will highlight some of these newest developments in heavy plates for bridge building. For example, for small span railway bridges the so-called thick plate trough bridges have proven to be a favourable concept. Very heavy plates with single plate weights up to 42 t allow building these bridges very efficiently out of one or very few single plates. Another interesting development is the so-called longitudinally profiled plates which allow a varying plate thickness along the actual loading profile. As last point the rising entry of higher strength steels in bridge building will be discussed and it will be shown why thermomechanically rolled plates are the ideal solution for these demands.

Heavy gauge plates for nuclear application

International Nuclear Information System (INIS)

Cheviet, A.; Roux, J.-H.

1977-01-01

The production of energy from nuclear sources leads to the building of very large vessels working under pressure at elevated temperatures, requiring very thick steel plate (from 50 mm to 300 mm). The plates necessary for the production of these vessels have to be as large as possible in order to reduce the length of welds on the vessels. Those two requirements lead to the manufacture of heavy products from 10 to 80 tons unit weight. These products are special, because their fabrication requires very big facilities and also extremely high quality of the steel. The main points are: high cleanliness; properties as homogeneous as possible. The tests carried out on industrially produced plates (especially on a plate of 200 mm thick show the level of quality that can be reached [fr

Experimental studies of oxidic molten corium-vessel steel interaction

International Nuclear Information System (INIS)

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

2001-01-01

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

Experimental studies of oxidic molten corium-vessel steel interaction

Energy Technology Data Exchange (ETDEWEB)

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

2001-12-01

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

Splitting in Dual-Phase 590 high strength steel plates

International Nuclear Information System (INIS)

Yang Min; Chao, Yuh J.; Li Xiaodong; Tan Jinzhu

2008-01-01

Charpy V-notch impact tests on 5.5 mm thick, hot-rolled Dual-Phase 590 (DP590) steel plate were evaluated at temperatures ranging from 90 deg. C to -120 deg. C. Similar tests on 2.0 mm thick DP590 HDGI steel plate were also conducted at room temperature. Splitting or secondary cracks was observed on the fractured surfaces. The mechanisms of the splitting were then investigated. Fracture surfaces were analyzed by optical microscope (OM) and scanning electron microscope (SEM). Composition of the steel plates was determined by electron probe microanalysis (EPMA). Micro Vickers hardness of the steel plates was also surveyed. Results show that splitting occurred on the main fractured surfaces of hot-rolled steel specimens at various testing temperatures. At temperatures above the ductile-brittle-transition-temperature (DBTT), -95 deg. C, where the fracture is predominantly ductile, the length and amount of splitting decreased with increasing temperature. At temperatures lower than the DBTT, where the fracture is predominantly brittle, both the length and width of the splitting are insignificant. Splitting in HDGI steel plates only appeared in specimens of T-L direction. The analysis revealed that splitting in hot-rolled plate is caused by silicate and carbide inclusions while splitting in HDGI plate results from strip microstructure due to its high content of manganese and low content of silicon. The micro Vickers hardness of either the inclusions or the strip microstructures is higher than that of the respective base steel

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

Effect of Al and N on the toughness of heavy section steel plates

International Nuclear Information System (INIS)

Kikutake, Tetsuo; Tokunaga, Yoshikuni; Nakao, Hitoji; Ito, Kametaro; Takaishi, Shogo.

1988-01-01

The effect of Al and N on the notch toughness and tensile strength of heavy section pressure vessel steel plates is investigated. Notch toughness of steel A533B (Mn-Mo-Ni), which has mixed microstructure of ferrite and bainite, is drastically changed by the ratio of sol.N/sol.Al. With metallurgical observations, it is revealed that AlN morphology is influenced by the ratio of sol.N/sol.Al through the level of solute Al(C Al ). At the heat treatment of heavy section steel plate, AlN shows OSTWALD ripening and its speed depends upon C Al . When Al is added (Al ≥ 0.010%) in steel and sol.N/sol.Al ≤ 0.5, C Al remains low. This prevents AlN ripening, and brings fine austenite grain size and high toughness. On the other hand, when sol.N/sol.Al Al becomes high and this gives poor toughness through coarse AlN precipitates and coarse austenite grain. Therefore, controll of sol.N/sol.Al over 0.5 is favorable to keep high toughness in A533B steel. In steel A387-22 (Cr-Mo) which has full bainitic microstructure, too fine austenite grain brings about poor hardenability, and polygonal ferrite, which brings about both poor strength and tughness, appears in microstructure. Then sol.N/sol.Al < 0.5 is better to give high hardenability in steel A387-22. (author)

The annal of british RPV steel plates for first nuclear power station in Japan (1). Unforseen accidents araised before nuclear power plant open

International Nuclear Information System (INIS)

Miyoshi, Shigeru

2011-01-01

This article described the author's experiences of reactor vessel steel plates for the first nuclear power station, Tokai-mura reactor. The station is an advanced Calder Hall type. The electrical output is 166 MWe. The reactor vessel was spherical with internal diameter of 189 cm and wall thickness of 83 mm. Material was a fine-grain, aluminum-killed steel. Each part of pressure vessel, bottom cap, belt 1, 2, 3, 4 and top cap, were prefabricated with welding of plates, then lifted into the reactor building and assembled with welding. Steel plates were imported from UK, press formed to spherical segments in Japan and transferred to the site. Ultrasonic testing, magnetic particle testing of groove face (crack detection), sizing of groove and sulfur print tests were performed as an on-site acceptance testing. Inclusions and lamination openings were observed at groove faces due to gas flame cutting. White spot was observed at rupture face of tensile test specimen. At the liquid penetration testing after back gauging of extra seam, a crack-like indication with length of less than 3 mm was observed. Reexamination of groove face by magnetic particles testing showed indications of inclusion cloud or alumina cloud. These would be cracks caused by hydrogen embrittlement. (T. Tanaka)

Electroless Plated Nanodiamond Coating for Stainless Steel Passivation

Energy Technology Data Exchange (ETDEWEB)

Li, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Korinko, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Spencer, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Stein, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-15

Tritium gas sample bottles and manifold components require passivation surface treatments to minimize the interaction of the hydrogen isotopes with surface contamination on the stainless steel containment materials. This document summarizes the effort to evaluate electroless plated nanodiamond coatings as a passivation layer for stainless steel. In this work, we developed an electroless nanodiamond (ND)-copper (Cu) coating process to deposit ND on stainless steel parts with the diamond loadings of 0%, 25% and 50% v/v in a Cu matrix. The coated Conflat Flanged Vessel Assemblies (CFVAs) were evaluated on surface morphology, composition, ND distribution, residual hydrogen release, and surface reactivity with deuterium. For as-received Cu and ND-Cu coated CFVAs, hydrogen off-gassing is rapid, and the off-gas rates of H₂ was one to two orders of magnitude higher than that for both untreated and electropolished stainless steel CFVAs, and hydrogen and deuterium reacted to form HD as well. These results indicated that residual H₂ was entrapped in the Cu and ND-Cu coated CFVAs during the coating process, and moisture was adsorbed on the surface, and ND and/or Cu might facilitate catalytic isotope exchange reaction for HD formation. However, hydrocarbons (i.e., CH₃) did not form, and did not appear to be an issue for the Cu and ND-Cu coated CFVAs. After vacuum heating, residual H₂ and adsorbed H₂O in the Cu and ND-Cu coated CFVAs were dramatically reduced. The H₂ off-gassing rate after the vacuum treatment of Cu and 50% ND-Cu coated CFVAs was on the level of 10^-14 l mbar/s cm², while H₂O off-gas rate was on the level of 10^-15 l mbar/s cm², consistent with the untreated or electropolished stainless steel CFVA, but the HD formation remained. The Restek EP bottle was used as a reference for this work. The Restek Electro-Polished (EP) bottle and their Sil

Magnetostrictive clad steel plates for high-performance vibration energy harvesting

Science.gov (United States)

Yang, Zhenjun; Nakajima, Kenya; Onodera, Ryuichi; Tayama, Tsuyoki; Chiba, Daiki; Narita, Fumio

2018-02-01

Energy harvesting technology is becoming increasingly important with the appearance of the Internet of things. In this study, a magnetostrictive clad steel plate for harvesting vibration energy was proposed. It comprises a cold-rolled FeCo alloy and cold-rolled steel joined together by thermal diffusion bonding. The performances of the magnetostrictive FeCo clad steel plate and conventional FeCo plate cantilevers were compared under bending vibration; the results indicated that the clad steel plate construct exhibits high voltage and power output compared to a single-plate construct. Finite element analysis of the cantilevers under bending provided insights into the magnetic features of a clad steel plate, which is crucial for its high performance. For comparison, the experimental results of a commercial piezoelectric bimorph cantilever were also reported. In addition, the cold-rolled FeCo and Ni alloys were joined by thermal diffusion bonding, which exhibited outstanding energy harvesting performance. The larger the plate volume, the more the energy generated. The results of this study indicated not only a promising application for the magnetostrictive FeCo clad steel plate as an efficient energy harvester, related to small vibrations, but also the notable feasibility for the formation of integrated units to support high-power trains, automobiles, and electric vehicles.

The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

Energy Technology Data Exchange (ETDEWEB)

Peacock, A., E-mail: alan.peacock@ipp.mpg.de [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Girlinger, A. [MAN Diesel and Turbo SE D-94469 Deggendorf (Germany); Vorkoeper, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 17491 Greifswald (Germany); Boscary, J.; Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hurd, F. [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)

2011-10-15

320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m{sup 2}, are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m{sup 2} and a maximum heat load of 200 kW/m{sup 2} with a water velocity of approximately 2 m s{sup -1}. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m{sup 2} for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

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

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

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

Flaw behavior in mechanically loaded clad plates

International Nuclear Information System (INIS)

Iskander, S.K.; Robinson, G.C.; Oland, C.B.

1989-01-01

A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

Corrosion of vessel steel during its interaction with molten corium

International Nuclear Information System (INIS)

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

2006-01-01

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

Low temperature radiation embrittlement for reactor vessel steels

International Nuclear Information System (INIS)

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

1978-01-01

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

Internal Friction of Pressure Vessel Steel Embrittlement

International Nuclear Information System (INIS)

Van Ouytsel, K.

2001-01-01

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

Progress in thermomechanical control of steel plates and their commercialization

Science.gov (United States)

Nishioka, Kiyoshi; Ichikawa, Kazutoshi

2012-01-01

The water-cooled thermomechanical control process (TMCP) is a technology for improving the strength and toughness of water-cooled steel plates, while allowing control of the microstructure, phase transformation and rolling. This review describes metallurgical aspects of the microalloying of steel, such as niobium addition, and discusses advantages of TMCP, for example, in terms of weldability, which is reduced upon alloying. Other covered topics include the development of equipment, distortions in steel plates, peripheral technologies such as steel making and casting, and theoretical modeling, as well as the history of property control in steel plate production and some early TMCP technologies. We provide some of the latest examples of applications of TMCP steel in various industries such as shipbuilding, offshore structures, building construction, bridges, pipelines, penstocks and cryogenic tanks. This review also introduces high heat-affected-zone toughness technologies, wherein the microstructure of steel is improved by the addition of fine particles of magnesium-containing sulfides and magnesium- or calcium-containing oxides. We demonstrate that thanks to ongoing developments TMCP has the potential to meet the ever-increasing demands of steel plates. PMID:27877477

Progress in thermomechanical control of steel plates and their commercialization

Directory of Open Access Journals (Sweden)

Kiyoshi Nishioka and Kazutoshi Ichikawa

2012-01-01

Full Text Available The water-cooled thermomechanical control process (TMCP is a technology for improving the strength and toughness of water-cooled steel plates, while allowing control of the microstructure, phase transformation and rolling. This review describes metallurgical aspects of the microalloying of steel, such as niobium addition, and discusses advantages of TMCP, for example, in terms of weldability, which is reduced upon alloying. Other covered topics include the development of equipment, distortions in steel plates, peripheral technologies such as steel making and casting, and theoretical modeling, as well as the history of property control in steel plate production and some early TMCP technologies. We provide some of the latest examples of applications of TMCP steel in various industries such as shipbuilding, offshore structures, building construction, bridges, pipelines, penstocks and cryogenic tanks. This review also introduces high heat-affected-zone toughness technologies, wherein the microstructure of steel is improved by the addition of fine particles of magnesium-containing sulfides and magnesium- or calcium-containing oxides. We demonstrate that thanks to ongoing developments TMCP has the potential to meet the ever-increasing demands of steel plates.

Thermal embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

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

1995-01-01

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

Characterization of four prestressed concrete reactor vessel liner steels

International Nuclear Information System (INIS)

Nanstad, R.K.

1980-12-01

A program of fracture toughness testing and analysis is being performed with PCRV steels for HTGRs. This report focuses on background information for the base materials and results of characterization testing, such as tensile and impact properties, chemical composition, and microstructural examination. The steels tested were an SA-508 class 1 forging, two plates of SA-537 class 1, and one plate of SA-537 class 2. Tensile requirements in effect at the time of procurement are met by all four steels. However, the SA-537 class 2 plate would not meet the minimum requirement for yield strength. Drop-weight and Charpy impact tests verified that the RT/sub NDT/ is equal to the NDT for each steel. Charpy impact energies at the NDT range from 40 J (30 ft-lb) for one heat of SA-537 class 1 to 100 J (74 ft-lb) for the SA-537 class 2 plate; upper-shelf energies range from 170 to 310 J (125 to 228 ft-lb) for the same two steels, respectively. The onset of upper-shelf energy occurred at temperatures ranging from 0 to 50 0 C

The prediction of failure of welded steel plates for pressure vessels

International Nuclear Information System (INIS)

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

1980-01-01

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

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

International Nuclear Information System (INIS)

Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

2004-01-01

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

Behavior of Equipment Support Beam Joint Directly Connected to A Steel-plate Concrete(SC) Wall

International Nuclear Information System (INIS)

Kim, K. S.; Kwon, K. J.

2008-01-01

To decrease the time for building nuclear power plants, a modular construction method, 'Steel-plate Concrete(SC)', has been investigated for over a decade. To construct a SC wall, a pair of steel plates are placed in parallel similar to a form-work in conventional reinforced concrete (RC) structures, and concrete is filled between the steel plates. Instead of removing the steel plates after the concrete has cured, the steel plates serve as components of the structural member. The exposed steel plate of SC structures serves as the base plate for the equipment support, and the headed studs welded to the steel plates are used as anchor bolts. Then, a support beam can be directly welded to the surface of the steel plate in any preferred position. In this study, we discuss the behavior and evaluation method of the equipment support joint directly connected to exposed steel plate of SC wall

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Fracture toughness of welded joints of ASTM A543 steel plate

International Nuclear Information System (INIS)

Susukida, H.; Uebayashi, T.; Yoshida, K.; Ando, Y.

1977-01-01

Fracture toughness and weldability tests have been performed on a high strength steel which is a modification of ASTM A543 Grade B Class 1 steel, with a view to using it for nuclear reactor containment vessels. The results showed that fracture toughness of welded joints of ASTM A543 modified high strength steel is superior and the steel is suitable for manufacturing the containment vessels

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

International Nuclear Information System (INIS)

Dunne, Druce; Li, Huijun; Jones, Christopher

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Research to sustain cases for Magnox-reactor steel pressure vessels

International Nuclear Information System (INIS)

Graham, W.J.

1997-01-01

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

Corrosion Behaviour of Nickel Plated Low Carbon Steel in Tomato Fluid

Directory of Open Access Journals (Sweden)

Oluleke OLUWOLE

2010-12-01

Full Text Available This research work investigated the corrosion resistance of nickel plated low carbon steel in tomato fluid. It simulated the effect of continuous use of the material in a tomato environment where corrosion products are left in place. Low carbon steel samples were nickel electroplated at 4V for 20, 25, 30 and 35 mins using Watts solution.The plated samples were then subjected to tomato fluid environment for for 30 days. The electrode potentials mV (SCE were measured every day. Weight loss was determined at intervals of 5 days for the duration of the exposure period. The result showed corrosion attack on the nickel- plated steel, the severity decreasing with the increasing weight of nickel coating on substrate. The result showed that thinly plated low carbon steel generally did not have any advantage over unplated steel. The pH of the tomato solution which initially was acidic was observed to progress to neutrality after 4 days and then became alkaline at the end of the thirty days test (because of corrosion product contamination of the tomatocontributing to the reduced corrosion rates in the plated samples after 10 days. Un-plated steel was found to be unsuitable for the fabrication of tomato processing machinery without some form of surface treatment - thick nickel plating is suitable as a protective coating in this environment.

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

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

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

Propagation of semi-elliptical surface cracks in ferritic and austenitic steel plates under thermal cyclic loading

International Nuclear Information System (INIS)

Bethge, K.

1989-05-01

Theoretical and experimental investigations of crack growth under thermal and thermomechanical fatigue loading are presented. The experiments were performed with a ferritic reactor pressure vessel steel 20 MnMoNi 5 5 and an austenitic stainless steel X6 CrNi 18 11. A plate containing a semi-elliptical surface crack is heated up to a homogeneous temperature and cyclically cooled down by a jet of cold water. On the basis of linear elastic fracture mechanics stress-intensity factors are calculated with the weight function method. The prediction of crack growth under thermal fatigue loading using data from mechanical fatigue tests is compared with the experimental result. (orig.) [de

Reactor pressure vessel steels

International Nuclear Information System (INIS)

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

1998-01-01

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

Manufacture of electron beam irradiation vessel and its characteristics

International Nuclear Information System (INIS)

Kanazawa, Takao; Haruyama, Yasuyuki; Yotsumoto, Keiichi

1992-05-01

Electron beam irradiation vessel, which is used for the irradiation of samples under an inert or a vacuum atmosphere, is made by considering the temperature control during or after irradiation. The vessel was composed of the temperature controlable samples supporting plate, beam slit with water cooling plate and the insert of thermosensor. The four samples supporting plate was produced with the materials made up of aluminium, stainless steel (SUS304), and copper. The stainless steel supporting plate has a heater inside the cooling pipes for the high temperature treatment of samples without exposure to atmosphere after the irradiation. In this report, the temperature distribution and dose characteristics such as dose distribution and effects of backscattered electron were studied by using several supporting plate and the comparison of the experimental results with the simulated results was also carried out. (author)

An Experimental Study on the Shear Hysteresis and Energy Dissipation of the Steel Frame with a Trapezoidal-Corrugated Steel Plate.

Science.gov (United States)

Shon, Sudeok; Yoo, Mina; Lee, Seungjae

2017-03-06

The steel frame reinforced with steel shear wall is a lateral load resisting system and has higher strength and shear performance than the concrete shear wall system. Especially, using corrugated steel plates in these shear wall systems improves out-of-plane stiffness and flexibility in the deformation along the corrugation. In this paper, a cyclic loading test of this steel frame reinforced with trapezoidal-corrugated steel plate was performed to evaluate the structural performance. The hysteresis behavior and the energy dissipation capacity of the steel frame were also compared according to the corrugated direction of the plate. For the test, one simple frame model without the wall and two frame models reinforced with the plate are considered and designed. The test results showed that the model reinforced with the corrugated steel plate had a greater accumulated energy dissipation capacity than the experimental result of the non-reinforced model. Furthermore, the energy dissipation curves of two reinforced frame models, which have different corrugated directions, produced similar results.

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Mechanical Behavior of BFRP-Steel Composite Plate under Axial Tension

Directory of Open Access Journals (Sweden)

Yunyu Li

2014-06-01

Full Text Available Combining the advantages of basalt fiber-reinforced polymer (BFRP material and steel material, a novel BFRP-steel composite plate (BSP is proposed, where a steel plate is sandwiched between two outer BFRP laminates. The main purpose of this research is to investigate the mechanical behavior of the proposed BSP under uniaxial tension and cyclic tension. Four groups of BSP specimens with four different BFRP layers and one control group of steel plate specimens were prepared. A uniaxial tensile test and a cyclic tensile test were conducted to determine the initial elastic modulus, postyield stiffness, yield strength, ultimate bearing capacity and residual deformation. Test results indicated that the stress-strain curve of the BSP specimen was bilinear prior to the fracture of the outer BFRP, and the BSP specimen had stable postyield stiffness and small residual deformation after the yielding of the inner steel plate. The postyield modulus of BSP specimens increased almost linearly with the increasing number of outer BFRP layers, as well as the ultimate bearing capacity. Moreover, the predicted results from the selected models under both monotonic tension and cyclic tension were in good agreement with the experimental data.

Blast response of corroded steel plates

Energy Technology Data Exchange (ETDEWEB)

Eslamimajd, Alireza; RahbarRanji, Ahmad [AmirKabir University of Technology, Tehran (Korea, Republic of)

2014-05-15

Numerical results for one- and both-sided corroded steel plates subjected to blast loading are presented. Finite element analysis, with ABAQUS software, is employed to determine the deformation and stress distributions. The results for the case of triangular pulse pressure on un-corroded plates are validated against literature-based data and then, detailed parametric studies are carried-out. The effects of influential parameters including, plate aspect ratio, degree of pit and different ratio of pit depth at each sides of the plate are investigated. The results show that position of pitted surface in respect to applied pressure is the most influential parameter on reduction of dynamic load carrying capacity of pitted plates. By increasing degree of pitting, reduction of dynamic load carrying capacity decrease more.

Fracture toughness of irradiated and recovered vessel steels

International Nuclear Information System (INIS)

Perosanz, F.; Lapena, J.

1998-01-01

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

Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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

Topic 1. Steels for light water reactor pressure vessels

International Nuclear Information System (INIS)

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

1980-01-01

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

Performance of Retrofitted Self-Compacting Concrete-Filled Steel Tube Beams Using External Steel Plates

Directory of Open Access Journals (Sweden)

Ahmed A. M. AL-Shaar

2018-01-01

Full Text Available Self-compacting concrete-filled steel tube (SCCFST beams, similar to other structural members, necessitate retrofitting for many causes. However, research on SCCFST beams externally retrofitted by bolted steel plates has seldom been explored in the literature. This paper aims at experimentally investigating the retrofitting performance of square self-compacting concrete-filled steel tube (SCCFST beams using bolted steel plates with three different retrofitting schemes including varied configurations and two different steel plate lengths under flexure. A total of 18 specimens which consist of 12 retrofitted SCCFST beams, three unretrofitted (control SCCFST beams, and three hollow steel tubes were used. The flexural behaviour of the retrofitted SCCFST beams was examined regarding flexural strength, failure modes, and moment versus deflection curves, energy absorption, and ductility. Experimental results revealed that the implemented retrofitting schemes efficiently improve the moment carrying capacity and stiffness of the retrofitted SCCFST beams compared to the control beams. The increment in flexural strength ranged from 1% to 46%. Furthermore, the adopted retrofitting schemes were able to restore the energy absorption and ductility of the damaged beams in the range of 35% to 75% of the original beam ductility. Furthermore, a theoretical model was suggested to predict the moment capacity of the retrofitted SCCFST beams. The theoretical model results were in good agreement with the test results.

Electroless nickel plating on stainless steels and aluminum

Science.gov (United States)

1966-01-01

Procedures for applying an adherent electroless nickel plating on 303 SE, 304, and 17-7 PH stainless steels, and 7075 aluminum alloy was developed. When heat treated, the electroless nickel plating provides a hard surface coating on a high strength, corrosion resistant substrate.

75 FR 81309 - Stainless Steel Plate from Belgium, Italy, Korea, South Africa, and Taiwan

Science.gov (United States)

2010-12-27

... (Second Review)] Stainless Steel Plate from Belgium, Italy, Korea, South Africa, and Taiwan AGENCY: United... on stainless steel plate from Belgium, Italy, Korea, South Africa, and Taiwan. SUMMARY: The... on stainless steel plate from Belgium, Italy, Korea, South Africa, and Taiwan would be likely to lead...

Effect of Plate Curvature on Blast Response of Structural Steel Plates

Science.gov (United States)

Veeredhi, Lakshmi Shireen Banu; Ramana Rao, N. V.; Veeredhi, Vasudeva Rao

2018-04-01

In the present work an attempt is made, through simulation studies, to determine the effect of plate curvature on the blast response of a door structure made of ASTM A515 grade 50 steel plates. A door structure with dimensions of 5.142 m × 2.56 m × 10 mm having six different radii of curvatures is analyzed which is subjected to blast load. The radii of curvature investigated are infinity (flat plate), 16.63, 10.81, 8.26, 6.61 and 5.56 m. In the present study, a stand-off distance of 11 m is considered for all the cases. Results showed that the door structure with smallest radius of curvature experienced least plastic deformation and yielding when compared to a door with larger radius of curvature with same projected area. From the present Investigation, it is observed that, as the radius of curvature of the plate increases, the deformation mode gradually shifts from indentation mode to flexural mode. The plates with infinity and 16.63 m radius of curvature have undergone flexural mode of deformation and plates with 6.61 and 5.56 m radius of curvature undergo indentation mode of deformation. Whereas, mixed mode of deformation that consists of both flexural and indentation mode of deformations are seen in the plates with radius of curvature 10.81 and 8.26 m. As the radius of curvature of the plate decreases the ability of the plate to mitigate the effect the blast loads increased. It is observed that the plate with smaller radius of curvature deflects most of the blast energy and results in least indentation mode of deformation. The most significant observation made in the present investigation is that the strain energy absorbed by the steel plate gets reduced to 1/3 rd when the radius of curvature is approximately equal to the stand-off distance which could be the critical radius of curvature.

Air-coupled ultrasonic through-transmission thickness measurements of steel plates.

Science.gov (United States)

Waag, Grunde; Hoff, Lars; Norli, Petter

2015-02-01

Non-destructive ultrasonic testing of steel structures provide valuable information in e.g. inspection of pipes, ships and offshore structures. In many practical applications, contact measurements are cumbersome or not possible, and air-coupled ultrasound can provide a solution. This paper presents air-coupled ultrasonic through-transmission measurements on a steel plate with thicknesses 10.15 mm; 10.0 mm; 9.8 mm. Ultrasound pulses were transmitted from a piezoelectric transducer at normal incidence, through the steel plate, and were received at the opposite side. The S1, A2 and A3 modes of the plate are excited, with resonance frequencies that depend on the material properties and the thickness of the plate. The results show that the resonances could be clearly identified after transmission through the steel plate, and that the frequencies of the resonances could be used to distinguish between the three plate thicknesses. The S1-mode resonance was observed to be shifted 10% down compared to a simple plane wave half-wave resonance model, while the A2 and S2 modes were found approximately at the corresponding plane-wave resonance frequencies. A model based on the angular spectrum method was used to predict the response of the through-transmission setup. This model included the finite aperture of the transmitter and receiver, and compressional and shear waves in the solid. The model predicts the frequencies of the observed modes of the plate to within 1%, including the down-shift of the S1-mode. Copyright © 2014 Elsevier B.V. All rights reserved.

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.

Experimental Study of Interactions Between Sub-oxidized Corium and Reactor Vessel Steel

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Granovsky, V.S.; Krushinov, E.V.; Vitol, S.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Fischer, M.; Piluso, P.; Altstadt, E.; Willschutz, H.G.; Fichoti, F.

2006-01-01

One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO 2 -ZrO 2 -Zr corium melt and VVER vessel steel was examined during the 2. Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and post-test analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090 deg. C. An empirical correlation for calculation of corrosion kinetics has been derived. (authors)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Immobilization of mesoporous silica particles on stainless steel plates

International Nuclear Information System (INIS)

Pasqua, Luigi; Morra, Marco

2017-01-01

A preliminary study aimed to the nano-engineering of stainless steel surface is presented. Aminopropyl-functionalized mesoporous silica is covalently and electrostatically anchored on the surface of stainless steel plates. The anchoring is carried out through the use of a nanometric spacer, and two different spacers are proposed (both below 2 nm in size). The first sample is obtained by anchoring to the stainless steel amino functionalized, a glutaryl dichloride spacer. This specie forms an amide linkage with the amino group while the unreacted acyl groups undergo hydrolysis giving a free carboxylic group. The so-obtained functionalized stainless steel plate is used as substrate for anchoring derivatized mesoporous silica particles. The second sample is prepared using 2-bromo-methyl propionic acid as spacer (BMPA). Successively, the carboxylic group of propionic acid is condensed to the aminopropyl derivatization on the external surface of the mesoporous silica particle through covalent bond. In both cases, a continuous deposition (coating thickness is around 10 μm) is obtained, in fact, XPS data do not reveal the metal elements constituting the plate. The nano-engineering of metal surfaces can represent an intriguing opportunity for producing long-term drug release or biomimetic surface.

Immobilization of mesoporous silica particles on stainless steel plates

Energy Technology Data Exchange (ETDEWEB)

Pasqua, Luigi, E-mail: luigi.pasqua@unical.it [University of Calabria, Department of Environmental and Chemical Engineering (Italy); Morra, Marco, E-mail: mmorra@nobilbio.com [Via Valcastellana 26 (Italy)

2017-03-15

A preliminary study aimed to the nano-engineering of stainless steel surface is presented. Aminopropyl-functionalized mesoporous silica is covalently and electrostatically anchored on the surface of stainless steel plates. The anchoring is carried out through the use of a nanometric spacer, and two different spacers are proposed (both below 2 nm in size). The first sample is obtained by anchoring to the stainless steel amino functionalized, a glutaryl dichloride spacer. This specie forms an amide linkage with the amino group while the unreacted acyl groups undergo hydrolysis giving a free carboxylic group. The so-obtained functionalized stainless steel plate is used as substrate for anchoring derivatized mesoporous silica particles. The second sample is prepared using 2-bromo-methyl propionic acid as spacer (BMPA). Successively, the carboxylic group of propionic acid is condensed to the aminopropyl derivatization on the external surface of the mesoporous silica particle through covalent bond. In both cases, a continuous deposition (coating thickness is around 10 μm) is obtained, in fact, XPS data do not reveal the metal elements constituting the plate. The nano-engineering of metal surfaces can represent an intriguing opportunity for producing long-term drug release or biomimetic surface.

Fabrication and mechanical test data for the four 6-inch-thick intermediate test vessels made from steel plate for the Heavy Section Steel Program

International Nuclear Information System (INIS)

Childress, C.E.

1976-01-01

The HSST Program has among its goals the objective of demonstrating the capability to predict safe behavior of thick-walled pressure vessels containing flaws of known dimensions under frangible, transitional, and tough loading regimes. To accomplish these objectives the program is conducting a series of tests involving 6-in.-thick pressure vessels which will serve as test specimens for assisting in the characterization of failure under these loading conditions. Among the vessels a number of parameters, such as weld type, weld location, flaw size and shape, and test temperature and pressure, will be selectively varied to show that a rationale exists for dealing with the varied stress and metallurgical states which normally exist in commercial nuclear reactor vessels. Each vessel will serve as a go, no-go determination of critical flaw size for a specific set of test parameters. Item 4 of the previous issues in this series covers the fabrication details of the first six 6-in.-thick test vessels, which were fabricated from ASTM A-508 Cl 2 forging materials. This report covers the fabrication details of four additional 6-in.-thick intermediate test vessels having shell courses fabricated from ASTM A-533 Gr B Cl 1 plate. The remaining components were made from forgings. Essentially this report is a continuation of ORNL-TM-4351; it describes the manufacturing details of the individual parts and their ultimate assembly into finished vessels. Details concerning chemical composition and mechanical and nondestructive test data are presented

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

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

2003-01-01

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

Ultrasonic stress evaluation through thickness of a stainless steel pressure vessel

International Nuclear Information System (INIS)

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

2014-01-01

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

Structure of steel reactor building and construction method therefor

International Nuclear Information System (INIS)

Yamakawa, Toshikimi.

1997-01-01

The building of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevation of inner pressure and keeping airtightness, and shielding concretes are filled between the double steel plate walls. It also has empty double steel plate walls not filled with concretes and has pipelines, vent ducts, wirings and a support structures for attaching them between the double steel plate walls. It is endurable to a great inner pressure satisfactory and keeps airtightness by the two spaced steel plates. It can be greatly reduced in the weight, and can be manufactured efficiently with high quality in a plant by so called module construction, and the dimension of the entire of the reactor building can be reduced. It is constructed in a dock, transported on the sea while having the space between the two steel plate walls as a ballast tanks, placed in the site, and shielding concretes are filled between the double steel plate walls. The term for the construction can be reduced, and the cost for the construction can be saved. (N.H.)

Application of low energy electron beam to precoated steel plates

International Nuclear Information System (INIS)

Koshiishi, Kenji

1989-01-01

Recently in the fields of home electric appliances, machinery and equipment and interior building materials, the needs for the precoated steel plates having the design and function of high class increase rapidly. In order to cope with such needs, the authors have advanced the examination on the application of electron beam hardening technology to precoated steel plates, and developed the precoated steel plates of high grade and high design 'Super Tecstar EB Series' by utilizing low energy electron beam. The features of this process are (1) hardening can be done at room temperature in a short time-thermally weak films can be adhered, (2) high energy irradiation-the hardening of thick enamel coating and the adhesion of colored films are feasible, (3) the use of monomers of low molecular weight-by high crosslinking, the performance of high sharpness, high hardness, anti-contamination property and so on can be given. The application to precoated steel plate production process is the coating and curing of electron beam hardening type paints, the coating of films with electron beam hardening type adhesives, and the reforming of surface polymer layers by impregnating monomers and causing graft polymerization with electron beam irradiation. The outline of the Super Tecstar EB Series is described. (K.I.)

Experimental Study on Temperature Behavior of SSC (Stiffened Steel Plate Concrete) Structures

International Nuclear Information System (INIS)

Lee, K. J.; Ham, K. W.; Park, D. S.; Kwon, K. J.

2008-01-01

SSC(Stiffened Steel plate Concrete) module method uses steel plate instead of reinforcing bar and mold in existing RC structure. Steel plate modules are fabricated in advance, installed and poured with concrete in construction field, so construction period is remarkably shortened by SC module technique. In case of existence of temperature gap between internal and external structure surface such as containment building, thermal stress is taken place and as a result of it, structural strength is deteriorated. In this study, we designed two test specimens and several tests with temperature heating were conducted to evaluate temperature behavior of SSC structures and RC structure

Fragmentation of armor piercing steel projectiles upon oblique perforation of steel plates

Directory of Open Access Journals (Sweden)

Aizik F.

2012-08-01

Full Text Available In this study, a constitutive strength and failure model for a steel core of a14.5 mm API projectile was developed. Dynamic response of a projectile steel core was described by the Johnson-Cook constitutive model combined with principal tensile stress spall model. In order to obtain the parameters required for numerical description of projectile core material behavior, a series of planar impact experiments was done. The parameters of the Johnson-Cook constitutive model were extracted by matching simulated and experimental velocity profiles of planar impact. A series of oblique ballistic experiments with x-ray monitoring was carried out to study the effect of obliquity angle and armor steel plate thickness on shattering behavior of the 14.5 mm API projectile. According to analysis of x-ray images the fragmentation level increases with both steel plate thickness and angle of inclination. The numerical modeling of the ballistic experiments was done using commercial finite element code, LS-DYNA. Dynamic response of high hardness (HH armor steel was described using a modified Johnson-Cook strength and failure model. A series of simulations with various values of maximal principal tensile stress was run in order to capture the overall fracture behavior of the projectile’s core. Reasonable agreement between simulated and x-ray failure pattern of projectile core has been observed.

Experimental study on behavior of RC panels covered with steel plates subjected to missile impact

International Nuclear Information System (INIS)

Jun Hashimoto; Katsuki Takiguchi; Koshiro Nishimura; Kazuyuki Matsuzawa; Mayuko Tsutsui; Yasuhiro Ohashi; Isao Kojima; Haruhiko Torita

2005-01-01

This paper describes an experimental study on the behavior of concrete panels with steel plate subjected to missile impact. Two tests were carried out, divided in accordance with the types of projectile, non-deformable and deformable. In all, 40 specimens of 750 mm square were prepared. The panel specimen was suspended vertically by two steel wire ropes to allow free movement after projectile impact, and was subjected to a projectile. As a result, it is confirmed that a RC panel with steel plate on its back side has higher impact resistance performance than a RC panel and that thickness of concrete panel, thickness of steel plate and the impact velocity of the projectile have a great effect on the failure modes of steel concrete panels. Moreover, based on the experimental results, the quantitative evaluation method for impact resistance performance of RC panels covered with steel plates is examined. The formula for perforation velocity of a half steel concrete panel, proposed in accordance with the bulging height, is effective to evaluate the impact resistance performance of RC panels with steel plates. (authors)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

Makioka, Minoru; Hirano, Hiromichi

1976-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1984-07-01

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

Mechanical System Analysis of C-Frame for Steel Plate Thickness Gauge

International Nuclear Information System (INIS)

Sri Sudadiyo

2007-01-01

Nuclear base instrument is not only applied in the area of research such as medical and agriculture sciences, but also in the field of industry especially for thickness gauge. To the present at the steel industry, the gauge that is applied to cut plate thickness using infra-red ray method, it cannot result in accurately data. To solve that case, it is developed a thickness gauge of steel plate by using gamma ray method that it is named C-Frame. This thickness gauge is hoped that it could control in cutting the steel plate by on-line, accurate, and safe, therefore, it could socialize the advanced technology in the nuclear field to support the production process in domestic industries (national industries). The present study yields the calculations of mechanical system of that C-Frame including structure, detector support, source container of radioisotope, and transmission system, be also computed by running Professional Microsoft Fortran Version 5.10, NISA-II program, and AutoCAD program. From the obtained results could be known that the design meets the requirement, so that could be employed properly to measure the thickness of plate in the steel industries. (author)

Integrity of Magnox reactor steel pressure vessels

International Nuclear Information System (INIS)

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

1992-01-01

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

Large inelastic deformation analysis of steel pressure vessels at high temperature

International Nuclear Information System (INIS)

Ikonen, K.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Corrosion of steel tendons used in prestressed concrete pressure vessels

International Nuclear Information System (INIS)

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

The purpose of this investigation was to determine the corrosion behavior of a high strength steel (ASTM A416-74 grade 270), typical of those used as tensioning tendons in prestressed concrete pressure vessels, in several corrosive environments and to demonstrate the protection afforded by coating the steel with either of two commercial petroleum-base greases or Portland Cement grout. In addition, the few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors are reviewed. The susceptibility of the steel to stress corrosion cracking and hydrogen embrittlement and its general corrosion rate were determined in several salt solutions. Wires coated with the greases and grout were soaked for long periods in the same solutions and changes in their mechanical properties were subsequently determined. All three coatings appeared to give essentially complete protection but small flaws in the grease coatings were detrimental; flaws or cracks less than 1 mm wide in the grout were without effect

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

James, L.A.

1977-01-01

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

Niobium Application, Metallurgy and Global Trends in Pressure Vessel Steels

Science.gov (United States)

Jansto, Steven G.

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

Free vibration analysis of corroded steel plates

Energy Technology Data Exchange (ETDEWEB)

Eslami-Majd, Alireza; Rahbar-Ranji, Ahmad [AmirKabir University of Technology, Tehran (Iran, Islamic Republic of)

2014-06-15

Vibration analysis of unstiffened/stiffened plates has long been studied due to its importance in the design and condition assessments of ship and offshore structures. Corrosion is inevitable in steel structures and has been so far considered in strength analysis of structures. We studied the free vibration of pitted corroded plates with simply supported boundary conditions. Finite element analysis, with ABAQUS, was used to determine the natural frequencies and mode shapes of corroded plates. Influential parameters including plate aspect ratio, degree of pit, one-sided/both-sided corroded plate, and different corrosion patterns were investigated. By increasing the degree of corrosion, reduction of natural frequency increases. Plate aspect ratio and plate dimensions have no influence on reduction of natural frequency. Different corrosion patterns on the surface of one-sided corroded plates have little influence on reduction of natural frequency. Ratio of pit depth over plate thickness has no influence on the reduction of natural frequency. The reduction of natural frequency in both-sided corroded plates is higher than one-sided corroded plates with the same amount of total corrosion loss. Mode shapes of vibration would change due to corrosion, except square mode shapes.

Corrosion of vessel steel during its interaction with molten corium

International Nuclear Information System (INIS)

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

2006-01-01

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

Corrosion of vessel steel during its interaction with molten corium

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-15

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation)], E-mail: bechta@sbor.spb.su; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almiashev, V.I. [Institute of Silicate Chemistry, Russian Academy of Sciences (ISCh RAS), St. Petersburg (Russian Federation); Lopukh, D.B. [SPb State Electrotechnical University (SPbGETU), St. Petersburg (Russian Federation); Bottomley, D. [EUROPAISCHE KOMMISSION, Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA/DEN/DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Rossendorf (FZR), Dresden (Germany); Fichot, F. [IRSN/DPAM/SEMCA, St. Paul lez Durance (France); Kymalainen, O. [FORTUM Nuclear Services Ltd., Espoo (Finland)

2009-06-15

The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

International Nuclear Information System (INIS)

Bechta, S.V.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

2009-01-01

The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

Energy Technology Data Exchange (ETDEWEB)

Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

2014-10-15

Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - Abstract: During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR)

Microstructural evolution in reactor pressure vessel steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

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

2000-09-01

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

Development and Technology of Large Thickness TMCP Steel Plate with 390MPA Grade Used for Engineering Machinery

Science.gov (United States)

Wang, Xiaoshu; Zhang, Zhijun; Zhang, Peng

Recently, with the rapid upgrading of the equipment in the steel Corp, the rolling technology of TMCP has been rapidly developed and widely applied. A large amount of steel plate has been produced by using the TMCP technology. The TMCP processes have been used more and more widely and replaced the heat treatment technology of normalizing, quenching and tempering heat process. In this paper, low financial input is considered in steel plate production and the composition of the steel has been designed with low C component, a limited alloy element of the Nb, and certain amounts of Mn element. During the continuous casting process, the size of the continuous casting slab section is 300 mm × 2400 mm. The rolling technology of TMCP is controlled at a lower rolling and red temperature to control the transformation of the microstructure. Four different rolling treatments are chosen to test its effects on the 390MPa grade low carbon steel of bainitic microstructure and properties. This test manages to produce a proper steel plate fulfilling the standard mechanical properties. Specifically, low carbon bainite is observed in the microstructure of the steel plate and the maximum thickness of steel plate under this TMCP technology is up to 80mm. The mechanical property of the steel plate is excellent and the KV2 at -40 °C performs more than 200 J. Moreover, the production costs are greatly reduced when the steel plate is produced by this TMCP technology when replacing the current production process of quenching and tempering. The low cost steel plate could well meet the requirements of producing engineering machinery in the steel market.

Variability of mechanical properties of nuclear pressure vessel steels

International Nuclear Information System (INIS)

Petrequin, P.; Soulat, P.

1980-01-01

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

Design Review Report for Concrete Cover Block Replaced by Steel Plate

Energy Technology Data Exchange (ETDEWEB)

JAKA, O.M.

2000-07-27

The design for the steel cover plates to replace concrete cover blocks for U-109 was reviewed and approved in a design review meeting. The design for steel plates to replace concrete blocks were reviewed and approved by comparison and similarity with U-109 for the following additional pits: 241-U-105. 241-I-103, 241-Ax-101. 241-A-101, 241-SX-105, 241-S-A, 241-S-C, 241-SX-A.

78 FR 4385 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From the Republic of Korea: Preliminary...

Science.gov (United States)

2013-01-22

...-Quality Steel Plate Products From the Republic of Korea: Preliminary Results of Antidumping Duty... the antidumping duty order on certain cut-to- length carbon-quality steel plate products (CTL plate... Carbon-Quality Steel Plate Products from the Republic of Korea'' dated concurrently with this notice...

Stålplader gav dobbelt bæreevne (Steel plates doubled the load bearing capacity)

DEFF Research Database (Denmark)

Nielsen, Jan Broch

1999-01-01

Abstract from an examination of motor road bridge beams reinforced with steel plates on the sides and bottom. The plates doubled the load bearing capacity of the beams.......Abstract from an examination of motor road bridge beams reinforced with steel plates on the sides and bottom. The plates doubled the load bearing capacity of the beams....

Steels and welding nuclear

International Nuclear Information System (INIS)

Sessa, M.; Milella, P.P.

1987-01-01

This ENEA Data-Base regards mechanical properties, chemical composition and heat treatments of nuclear pressure vessel materials: type A533-B, A302-B, A508 steel plates and forgings, submerged arc welds and HAZ before and after nuclear irradiation. Irradiation experiments were generally performed in high flux material test reactors. Data were collected from international available literature about water nuclear reactors pressure vessel materials embrittlement

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Evaluation of Steel Shear Walls Behavior with Sinusoidal and Trapezoidal Corrugated Plates

Directory of Open Access Journals (Sweden)

Emad Hosseinpour

2015-01-01

Full Text Available Reinforcement of structures aims to control the input energy of unnatural and natural forces. In the past four decades, steel shear walls are utilized in huge constructions in some seismic countries such as Japan, United States, and Canada to lessen the risk of destructive forces. The steel shear walls are divided into two types: unstiffened and stiffened. In the former, a series of plates (sinusoidal and trapezoidal corrugated with light thickness are used that have the postbuckling field property under overall buckling. In the latter, steel profile belt series are employed as stiffeners with different arrangement: horizontal, vertical, or diagonal in one side or both sides of wall. In the unstiffened walls, increasing the thickness causes an increase in the wall capacity under large forces in tall structures. In the stiffened walls, joining the stiffeners to the wall is costly and time consuming. The ANSYS software was used to analyze the different models of unstiffened one-story steel walls with sinusoidal and trapezoidal corrugated plates under lateral load. The obtained results demonstrated that, in the walls with the same dimensions, the trapezoidal corrugated plates showed higher ductility and ultimate bearing compared to the sinusoidal corrugated plates.

Development of PIE techniques for irradiated LWR pressure vessel steels

International Nuclear Information System (INIS)

Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

1999-01-01

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

Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

International Nuclear Information System (INIS)

Hennion, A.

1999-03-01

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

Product consistency testing of three reference glasses in stainless steel and perfluoroalkoxy resin vessels

International Nuclear Information System (INIS)

Olson, K.M.; Smith, G.L.; Marschman, S.C.

1995-03-01

Because of their chemical durability, silicate glasses have been proposed and researched since the mid-1950s as a medium for incorporating high-level radioactive waste (HLW) generated from processing of nuclear materials. A number of different waste forms were evaluated and ranked in the early 1980s; durability (leach resistance) was the highest weighted factor. Borosilicate glass was rated the best waste form available for incorporation of HLW. Four different types of vessels and three different glasses were used to study the possible effect of vessel composition on durability test results from the Production Consistency Test (PCT). The vessels were 45-m 304 stainless steel vessels, 150-m 304 L stainless steel vessels, and 60-m perfluoroalkoxy (PFA) fluoropolymer resin vessels. The three glasses were the Environmental Assessment glass manufactured by Corning Incorporated and supplied by Westinghouse Savannah River company, and West Valley Nuclear Services reference glasses 5 and 6, manufactured and supplied by Catholic University of America. Within experimental error, no differences were found in durability test results using the 3 different glasses in the 304L stainless steel or PFA fluoropolymer resin vessels over the seven-day test period

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Nuclear reactors sited deep underground in steel containment vessels

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

Laser cutting of thick steel plates with 30 kW fiber laser for nuclear decommissioning

International Nuclear Information System (INIS)

Tamura, Koji

2015-01-01

Laser cutting technologies of the thick steel plates for the nuclear decommissioning were developed with a 30 kW fiber laser. Plates of stainless steel and carbon steel more than 100 mm thick were successfully cut, indicating that this technology is promising for the application to the nuclear decommissioning. (author)

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

International Nuclear Information System (INIS)

Vacek, N.; Stoces, B.

1980-01-01

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

Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

Design and analysis of reactor containment of steel-concrete composite laminated shell

International Nuclear Information System (INIS)

Ichikawa, K.

1977-01-01

Reinforced and prestressed concrete containments for reactors have been developed in order to avoid the difficulties of welding of steel containments encountered as their capacities have become large: growing thickness of steel shells gave rise to the requirement of stress relief at the construction sites. However, these concrete vessels also seem to face another difficulty: the lack of shearing resistance capacity. In order to improve the shearing resistance capacity of the containment vessel, while avoiding the difficulty of welding, a new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented. The results of model tests in 1:30 scale are also reported. (Auth.)

Numerical Study on Ultimate Behaviour of Bolted End-Plate Steel Connections

Directory of Open Access Journals (Sweden)

R.E.S. Ismail

Full Text Available Abstract Bolted end-plate steel connections have become more popular due to ease of fabrication. This paper presents a three dimension Finite Element Model (FEM, using the multi-purpose software ABAQUS, to study the effect of different geometrical parameters on the ultimate behavior of the connection. The proposed model takes into account material and geometrical non-linearities, initial imperfection, contact between adjacent surfaces and the pretension force in the bolts. The Finite Element results are calibrated with published experimental results ''briefly reviewed in this paper'' and verified that the numerical model can simulate and analyze the overall and detailed behavior of different types of bolted end-plate steel connections. Using verified FEM, parametric study is then carried out to study the ultimate behavior with variations in: bolt diameter, end-plate thickness, length of column stiffener, angle of rib stiffener. The results are examined with respect to the failure modes, the evolution of the resistance, the initial stiffness, and the rotation capacity. Finally, the ultimate behavior of the bolted end-plate steel connection is discussed in detail, and recommendations for the design purpose are made.

75 FR 29976 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From Italy: Extension of the Final...

Science.gov (United States)

2010-05-28

... DEPARTMENT OF COMMERCE International Trade Administration [A-475-826] Certain Cut-to-Length Carbon-Quality Steel Plate Products From Italy: Extension of the Final Results of Antidumping Duty Administrative...-quality steel plate products from Italy. See Certain Cut-to-Length Carbon-Quality Steel Plate Products...

Study on the application of thickened welds without post weld heat treatment for containment vessels

International Nuclear Information System (INIS)

Takeuchi, T.; Fukaya, T.; Sato, M.; Takano, G.

1978-01-01

As material for containment vessels, SGV49 steel plates are mainly used. However, those used for this purpose are limited in thickness to smaller than 38 mm. This is because the present standard requires welds thicker than 38 mm to be subjected to post weld heat treatment but operation on the site is practically difficult. In the case of 3-loop containment vessels of pressurized water type reactors, use of 38 mm SGV49 brings an increase in their height and this is disadvantageous from a seismic viewpoint. Therefore, use of 45 mm-thick steel material has become necessary in order to increase design internal pressure and reduce the height of the vessels. To investigate the propriety of the use of 45 mm-thick SGV49 for this purpose without post weld heat treatment we investigated the basic performances of base metal and welded joints. We also conducted large-scale embrittlement fracture tests (CT test, deep notch test, wide plate tensile test and ESSO test) in order to examine whether welds not subjected to post weld heat treatment are safe against embrittlement fracture under the operating conditions of the vessels. The results proved that the welds of SGV49 steel plates are safe enough under the operating conditions. (author)

Thermo-mechanical behaviour of FBTR reactor vessel due to natural convection in cover gas space

International Nuclear Information System (INIS)

Srinivasan, G.; Varadarajan, S.; Kapoor, R.P.

1988-01-01

Fast Breeder Test Reactor is a 40 MW(t), loop type sodium cooled reactor, similar in design to Rapsodie. The Reactor Assembly, which is the heart of FBTR, comprises the Reactor Vessel (RV) housed in a safety vessel within a concrete cell (A1 Cell). The RV which supports the core is shielded at the top by two rotatable plugs which are stacked with layers of borated graphite and steel. The smaller plug (SRP), is mounted excentric to the larger one (LRP). A nominal annular gap of 16 mm is provided between RV and LRP and between LRP and SRP to enable free rotation of the plugs. Stainless Steel insulation is fixed inside the steel vessel, to avoid overheating of the A1 Cell concrete. The core is supported by the Grid Plate (GP), bolted to the RV. During preheating, sodium charging and isothermal runs upto 350 0 C, temperature asymmetries were noticed in the reactor vessel wall in the cover gas space. This was attributable to convection currents in the annulus between RV and LRP. The asymmetries also resulted in a lateral shift of the grid plate. This paper discusses our experience in suppressing these convection currents, and minimising the grid plate shift

75 FR 59744 - Stainless Steel Plate From Belgium, Italy, Korea, South Africa, and Taiwan

Science.gov (United States)

2010-09-28

... (Second Review)] Stainless Steel Plate From Belgium, Italy, Korea, South Africa, and Taiwan AGENCY: United..., and Taiwan. SUMMARY: The Commission hereby gives notice that it will proceed with full reviews... antidumping duty orders on stainless steel plate from Belgium, Italy, Korea, South Africa, and Taiwan would be...

Experimental observations and modelling of thermal history within a steel plate during water jet impingement

International Nuclear Information System (INIS)

Liu, Z.D.; Fraser, D.; Samarasekera, I.V.; Lockhart, G.T.

2002-01-01

In order to investigate heat transfer of steel plates under a water jet impingement and to further simulate runout table operation in a hot strip mill, a full-scale pilot runout table facility was designed and constructed at the University of British Columbia (UBC). This paper describes the experimental details, data acquisition and data handling techniques for steel plates during water jet impingement by one circular water jet from an industrial header. Recorded visual observations at the impinging surface were obtained. The effects of cooling water temperature and impingement velocity on the heat transfer from a steel plate were studied. A two-dimensional finite element method-based transient inverse heat conduction model was developed. With the help of the model, heat fluxes and heat transfer coefficients along the impinging surface under various cooling conditions were calculated. The microstructural evolution of the steel plate was also investigated for the varying cooling conditions. Samples were obtained from each plate, polished, etched and then photographed. (author)

Seismic behavior and design of a primary shield structure consisting of steel-plate composite (SC) walls

Energy Technology Data Exchange (ETDEWEB)

Booth, Peter N., E-mail: boothpn@purdue.edu [Lyles School of Civil Engineering, Purdue University, W. Lafayette, IN (United States); Varma, Amit H., E-mail: ahvarma@purdue.edu [Lyles School of Civil Engineering, Purdue University, W. Lafayette, IN (United States); Sener, Kadir C., E-mail: ksener@purdue.edu [Lyles School of Civil Engineering, Purdue University, W. Lafayette, IN (United States); Mori, Kentaro, E-mail: kentaro_mori@mhi.co.jp [Mitsubishi Heavy Industries, Ltd, Kobe (Japan)

2015-12-15

This paper presents an analytical evaluation of the seismic behavior and design of a unique primary shield (PSW) structure consisting of steel-plate composite (SC) walls designed for a typical pressurized water reactor (PWR) nuclear power plant. Researchers in Japan have previously conducted a reduced (1/6th) scale test of a PSW structure to evaluate its seismic (lateral) load-deformation behavior. This paper presents the development and benchmarking of a detailed 3D nonlinear inelastic finite element (NIFE) model to predict the lateral load-deformation response and behavior of the 1/6th scale test structure. The PSW structure consists of thick SC wall segments with complex and irregular geometry that surround the central reactor vessel cavity. The wall segments have three layers of steel plates (one each on the interior and exterior surfaces and one embedded in the middle) that are anchored to the concrete infill with stud anchors. The results from the 3D NIFE analyses include: (i) the lateral load-deformation behavior of the PSW structure, (ii) the progression of yielding in the steel plates, concrete cracking, formation of compression struts, and (iii) the final failure mode. These results are compared and benchmarked using experimental measurements and observations reported by Shodo et al. (2003). The analytical results provide significant insight into the lateral behavior and strength of the PSW structure, and are used for developing a design approach. This design approach starts with ACI 349 code equations for reinforced concrete shear walls and modifies them for application to the PSW structure. A simplified 3D linear elastic finite element (LEFE) model of the PSW structure is also proposed as a conventional structural analysis tool for estimating the design force demands for various load combinations.

Seismic behavior and design of a primary shield structure consisting of steel-plate composite (SC) walls

International Nuclear Information System (INIS)

Booth, Peter N.; Varma, Amit H.; Sener, Kadir C.; Mori, Kentaro

2015-01-01

This paper presents an analytical evaluation of the seismic behavior and design of a unique primary shield (PSW) structure consisting of steel-plate composite (SC) walls designed for a typical pressurized water reactor (PWR) nuclear power plant. Researchers in Japan have previously conducted a reduced (1/6th) scale test of a PSW structure to evaluate its seismic (lateral) load-deformation behavior. This paper presents the development and benchmarking of a detailed 3D nonlinear inelastic finite element (NIFE) model to predict the lateral load-deformation response and behavior of the 1/6th scale test structure. The PSW structure consists of thick SC wall segments with complex and irregular geometry that surround the central reactor vessel cavity. The wall segments have three layers of steel plates (one each on the interior and exterior surfaces and one embedded in the middle) that are anchored to the concrete infill with stud anchors. The results from the 3D NIFE analyses include: (i) the lateral load-deformation behavior of the PSW structure, (ii) the progression of yielding in the steel plates, concrete cracking, formation of compression struts, and (iii) the final failure mode. These results are compared and benchmarked using experimental measurements and observations reported by Shodo et al. (2003). The analytical results provide significant insight into the lateral behavior and strength of the PSW structure, and are used for developing a design approach. This design approach starts with ACI 349 code equations for reinforced concrete shear walls and modifies them for application to the PSW structure. A simplified 3D linear elastic finite element (LEFE) model of the PSW structure is also proposed as a conventional structural analysis tool for estimating the design force demands for various load combinations.

Effect of heterogeneities on the thermoelectric power of pressure vessel steel

International Nuclear Information System (INIS)

Simonet, L.

2006-12-01

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

Cytotoxicity difference of 316L stainless steel and titanium reconstruction plate

Directory of Open Access Journals (Sweden)

Ni Putu Mira Sumarta

2011-03-01

Full Text Available Background: Pure titanium is the most biocompatible material today and used as a gold standard for metallic implants. However, stainless steel is still being used as implants because of its strength, ductility, lower price, corrosion resistant and biocompatibility. Purpose: This study was done to revealed the cytotoxicity difference between reconstruction plate made of 316L stainless steel and of commercially pure (CP titanium in baby hamster kidney-21 (BHK-21 fibroblast culture through MTT assay. Methods: Eight samples were prepared from reconstruction plates made of stainless steel type 316L grade 2 (Coen’s reconstruction plate® that had been cut into cylindrical form of 2 mm in diameter and 3 mm long. The other one were made of CP titanium (STEMA Gmbh® of 2 mm in diameter and 2,2 mm long; and had been cleaned with silica paper and ultrasonic cleaner, and sterilized in autoclave at 121° C for 20 minutes.9 Both samples were bathed into microplate well containing 50 μl of fibroblast cells with 2 x 105 density in Rosewell Park Memorial Institute-1640 (RPMI-1640 media, spinned at 30 rpm for 5 minutes. Microplate well was incubated for 24 and 48 hours in 37° C. After 24 hours, each well that will be read at 24 hour were added with 50 μl solution containing 5mg/ml MTT reagent in phosphate buffer saline (PBS solutions, then reincubated for 4 hours in CO2 10% and 37° C. Colorometric assay with MTT was used to evaluate viability of the cells population after 24 hours. Then, each well were added with 50 μl dimethyl sulfoxide (DMSO and reincubated for 5 minutes in 37° C. the wells were read using Elisa reader in 620 nm wave length. Same steps were done for the wells that will be read in 48 hours. Each data were tabulated and analyzed using independent T-test with significance of 5%. Results: This study showed that the percentage of living fibroblast after exposure to 316L stainless steel reconstruction plate was 61.58% after 24 hours and 62

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

International Nuclear Information System (INIS)

Porter, V.L.

1993-01-01

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

78 FR 29113 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From the Republic of Korea: Final...

Science.gov (United States)

2013-05-17

...-Quality Steel Plate Products From the Republic of Korea: Final Results of Antidumping Duty Administrative... administrative review of the antidumping duty order on certain cut-to-length carbon-quality steel plate products... duty order on certain cut-to-length carbon-quality steel plate products from the Republic of Korea...

The Development and Microstructure Analysis of High Strength Steel Plate NVE36 for Large Heat Input Welding

Science.gov (United States)

Peng, Zhang; Liangfa, Xie; Ming, Wei; Jianli, Li

In the shipbuilding industry, the welding efficiency of the ship plate not only has a great effect on the construction cost of the ship, but also affects the construction speed and determines the delivery cycle. The steel plate used for large heat input welding was developed sufficiently. In this paper, the composition of the steel with a small amount of Nb, Ti and large amount of Mn had been designed in micro-alloyed route. The content of C and the carbon equivalent were also designed to a low level. The technology of oxide metallurgy was used during the smelting process of the steel. The rolling technology of TMCP was controlled at a low rolling temperature and ultra-fast cooling technology was used, for the purpose of controlling the transformation of the microstructure. The microstructure of the steel plate was controlled to be the mixed microstructure of low carbon bainite and ferrite. Large amount of oxide particles dispersed in the microstructure of steel, which had a positive effects on the mechanical property and welding performance of the steel. The mechanical property of the steel plate was excellent and the value of longitudinal Akv at -60 °C is more than 200 J. The toughness of WM and HAZ were excellent after the steel plate was welded with a large heat input of 100-250 kJ/cm. The steel plate processed by mentioned above can meet the requirement of large heat input welding.

Magnetic and electrical properties of ITER vacuum vessel steels

International Nuclear Information System (INIS)

Mergia, K.; Apostolopoulos, G.; Gjoka, M.; Niarchos, D.

2007-01-01

Full text of publication follows: Ferritic steel AISI 430 is a candidate material for the lTER vacuum vessel which will be used to limit the ripple in the toroidal magnetic field. The magnetic and electrical properties and their temperature dependence in the temperature range 300 - 900 K of AISI 430 ferritic stainless steels are presented. The temperature variation of the coercive field, remanence and saturation magnetization as well as electrical resistivity and the effect of annealing on these properties is discussed. (authors)

Irradiation effects in strain aged pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-02-15

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

Design proposal for ultimate shear strength of tapered steel plate girders

Directory of Open Access Journals (Sweden)

A. Bedynek

2017-03-01

Full Text Available Numerous experimental and numerical studies on prismatic plate girders subjected to shear can be found in the literature. However, the real structures are frequently designed as non-uniform structural elements. The main objective of the research is the development of a new proposal for the calculation of the ultimate shear resistance of tapered steel plate girders taking into account the specific behaviour of such members. A new mechanical model is presented in the paper and it is used to show the differences between the behaviour of uniform and tapered web panels subjected to shear. EN 1993-1-5 design specifications for the determination of the shear strength for rectangular plates are improved in order to assess the shear strength of tapered elements. Numerical studies carried out on tapered steel plate girders subjected to shear lead to confirm the suitability of the mechanical model and the proposed design expression.

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

International Nuclear Information System (INIS)

Porter, V.L.

1994-01-01

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

Kinetics of annealing of irradiated surveillance pressure vessel steel

International Nuclear Information System (INIS)

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

1982-01-01

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

Micromechanisms of ductile stable crack growth in nuclear pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

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

1981-10-01

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

Residual stresses in a weldment of pressure vessel steel

International Nuclear Information System (INIS)

Gott, K.E.

1978-01-01

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

Seismic Performance and Design of Steel Plate Shear Walls with Low Yield Point Steel Infill Plates

OpenAIRE

Zirakian, Tadeh

2013-01-01

Steel plate shear walls (SPSWs) have been frequently used as the primary or part of the primary lateral force-resisting system in design of low-, medium-, and high-rise buildings. Their application has been based on two different design philosophies as well as detailing strategies. Stiffened and/or stocky-web SPSWs with improved buckling stability and high seismic performance have been mostly used in Japan, which is one of the pioneering countries in design and application of these systems. U...

Thermal Stress Cracking of Slide-Gate Plates in Steel Continuous Casting

Science.gov (United States)

Lee, Hyoung-Jun; Thomas, Brian G.; Kim, Seon-Hyo

2016-04-01

The slide-gate plates in a cassette assembly control the steel flow through the tundish nozzle, and may experience through-thickness cracks, caused by thermal expansion and/or mechanical constraint, leading to air aspiration and safety concerns. Different mechanisms for common and rare crack formation are investigated with the aid of a three-dimensional finite-element model of thermal mechanical behavior of the slide-gate plate assembly during bolt pretensioning, preheating, tundish filling, casting, and cooling stages. The model was validated with previous plant temperature measurements of a ladle plate during preheating and casting, and then applied to a typical tundish-nozzle slide-gate assembly. The formation mechanisms of different types of cracks in the slide-gate plates are investigated using the model and evaluated with actual slide-gate plates at POSCO. Common through-thickness radial cracks, found in every plate, are caused during casting by high tensile stress on the outside surfaces of the plates, due to internal thermal expansion. In the upper plate, these cracks may also arise during preheating or tundish filling. Excessive bolt tightening, combined with thermal expansion during casting may cause rare radial cracks in the upper and lower plates. Rare radial and transverse cracks in middle plate appear to be caused during tundish filling by impingement of molten steel on the middle of the middle plate that generates tensile stress in the surrounding refractory. The mechanical properties of the refractory, the bolt tightening conditions, and the cassette/plate design are all important to service life.

Results of steel containment vessel model test

International Nuclear Information System (INIS)

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

1998-05-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Pressure Vessel Steel Research: Belgian Activities

International Nuclear Information System (INIS)

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

1994-03-01

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

Pressure Vessel Steel Research: Belgian Activities

Energy Technology Data Exchange (ETDEWEB)

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

1994-03-01

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

Interaction between molten corium UO2+x-ZrO2-FeOy and VVER vessel steel

International Nuclear Information System (INIS)

Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A.; Gusarov, V. V.; Almiashev, V. I.; Lopukh, D. B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

2010-01-01

In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO 2+x -ZrO 2 -FeO y melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-01

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

Development and Application of TMCP Steel Plate in Coal Mining Machinery

Science.gov (United States)

Yongqing, Zhang; Liandeng, Yao; aimin, Guo; Sixin, Zhao; Guofa, Wang

Coal, as the most major energy in China, accounted for about 70% of China's primary energy production and consumption. While the percentage of coal as the primary energy mix would drop in the future due to serious smog pollution partly resulted from coal-burning, the market demand of coal will maintain because the progressive process of urbanization. In order to improve productivity and simultaneously decrease safety accidents, fully-mechanized underground mining technology based on complete equipment of powered support, armored face conveyor, shearer, belt conveyor and road-header have obtained quick development in recent years, of which powered support made of high strength steel plate accounts for 65 percent of total equipment investment, so, the integrated mechanical properties, in particular strength level and weldability, have a significant effects on working service life and productivity. Take hydraulic powered supports as example, this paper places priority to introduce the latest development of high strength steel plates of Q550, Q690 and Q890 for powered supports, as well as metallurgical design conception and production cost-benefits analysis between QT plate and TMCP plate. Through production and application practice, TMCP or DQ plate demonstrate great economic advantages compared with traditional QT plate.

The response of pressure vessel steel specimens on drop weight loading

International Nuclear Information System (INIS)

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

1979-01-01

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

Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2

International Nuclear Information System (INIS)

Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio

1997-01-01

A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)

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

International Nuclear Information System (INIS)

Chi, Byung Ha; Kim, Jeong Tae

2000-01-01

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

Improving electron beam weldability of heavy steel plates for PWR-steam generator

International Nuclear Information System (INIS)

Tomita, Yukio; Mabuchi, Hidesato; Koyama, Kunio

1996-01-01

Installation and replacement of many PWR-steam generators are planned inside and outside Japan. The steel plates for steam generators are heavy in thickness, and increase the number of welding passes and prolong the welding time. Electron beam welding (EBW) can greatly reduce the welding period compared with conventional welding methods (narrow-gap gas metal arc welding (GMAW) and submerged arc welding (SAW)). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Defect-free welding procedures were successfully established even in thick steel plates. The factors that deteriorate weld-metal (WM) toughness of EBW were investigated. The manufacturing process, which utilizes a new secondary refining process at steelmaking and a high-torque mill at plate mill in actual mass-production, were established. EBW base metal and WM have better properties including fracture toughness than those of conventional welding processes. As a result, an application of EBW to the fabrication of PWR-steam generators has become possible. Large amounts of ASTM A533 Gr B Cl 2 (JIS SQV2B) steel plates in actual PWR-steam generators have come to be produced (more than 1,500 ton) by applying EBW. (author)

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Sensitivity and statistical analysis within the elaboration of steel plated girder resistance

Czech Academy of Sciences Publication Activity Database

Melcher, J.; Škaloud, Miroslav; Kala, Z.; Karmazínová, M.

2009-01-01

Roč. 5, č. 2 (2009), s. 120-126 ISSN 1816-112X. [International conf. on steel and aluminium structures /6./. Oxford, 24.06.2007-27.06.2007] Institutional research plan: CEZ:AV0Z20710524 Keywords : steel structures * fatigue * sensitivity * imperfection * plated girder Subject RIV: JM - Building Engineering

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

International Nuclear Information System (INIS)

Sohn, Hee Dong

2012-02-01

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong

2012-02-15

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

77 FR 21527 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From the Republic of Korea: Final...

Science.gov (United States)

2012-04-10

... from the Republic of Korea. The review covers one manufacturer/ exporter. The period of review is...-Quality Steel Plate Products From the Republic of Korea: Final Results of Antidumping Duty Administrative... duty order on certain cut-to-length carbon-quality steel plate products (CTL plate) from the Republic...

Magnetic Barkhausen noise and magneto acoustic emission in pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Magnetic Barkhausen noise and magneto acoustic emission in pressure vessel steel

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

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

1976-08-01

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

Centralized Gap Clearance Control for Maglev Based Steel-Plate Conveyance System

Directory of Open Access Journals (Sweden)

GUNEY, O. F.

2017-08-01

Full Text Available The conveyance of steel-plates is one of the potential uses of the magnetic levitation technology in industry. However, the electromagnetic levitation systems inherently show nonlinear feature and are unstable without an active control. Well-known U-shaped or E-shaped electromagnets cannot provide redundant levitation with multiple degrees of freedom. In this paper, to achieve the full redundant levitation of the steel plate, a quadruple configuration of U shaped electromagnets has been proposed. To resolve the issue of instability and attain more robust levitation, a centralized control algorithm based on a modified PID controller (I PD is designed for each degree of freedom by using the Manabe canonical polynomial technique. The model of the system is carried out using electromechanical energy conversion princi¬ples and verified by 3-D FEM analysis. An experimental bench is built up to test the system performance under trajectory tracking and external disturbance excitation. The results confirm the effectiveness of the proposed system and the control approach to obtain a full redundant levitation even in case of disturbances. The paper demonstrates the feasibility of the con¬veyance of steel plates by using the quadruple configuration of U-shaped electromagnets and shows the merits of I-PD controller both in stabilization and increased robust levitation.

Nanosized TiN-SBR hybrid coating of stainless steel as bipolar plates for polymer electrolyte membrane fuel cells

International Nuclear Information System (INIS)

Kumagai, Masanobu; Myung, Seung-Taek; Asaishi, Ryo; Sun, Yang-Kook; Yashiro, Hitoshi

2008-01-01

In attempt to improve interfacial electrical conductivity of stainless steel for bipolar plates of polymer electrolyte membrane fuel cells, TiN nanoparticles were electrophoretically deposited on the surface of stainless steel with elastic styrene butadiene rubber (SBR) particles. From transmission electron microscopic observation, it was found that the TiN nanoparticles (ca. 50 nm) surrounded the spherical SBR particles (ca. 300-600 nm), forming agglomerates. They were well adhered on the surface of the type 310S stainless steel. With help of elasticity of SBR, the agglomerates were well fitted into the interfacial gap between gas diffusion layer (GDL) and stainless steel bipolar plate, and the interfacial contact resistance (ICR), simultaneously, was successfully reduced. A single cell using the TiN nanoparticles-coated bipolar plates, consequently, showed comparable cell performance with the graphite employing cell at a current density of 0.5 A cm -2 (12.5 A). Inexpensive TiN nanoparticle-coated type 310S stainless steel bipolar plates would become a possible alternate for the expensive graphite bipolar plates as use in fuel cell applications

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

Science.gov (United States)

Fekete, Balazs; Trampus, Peter

2015-09-01

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

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

International Nuclear Information System (INIS)

Witt, F.J.

1983-01-01

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

Heissdampfreaktor (HDR) steel-containment-vessel and floodwater-storage-tank structural-dynamics tests

International Nuclear Information System (INIS)

Arendts, J.G.

1982-01-01

Inertance (vibration) testing of two significant vessels at the Heissdampfreaktor (HDR) facility, located near Kahl, West Germany, was recently completed. Transfer functions were obtained for determination of the modal properties (frequencies, mode shapes and damping) of the vessels using two different test methods for comparative purposes. One of the vessels tested was the steel containment vessel (SCV). The SCV is approximately 180 feet high and 65 feet in diameter with a 1.2-inch wall thickness. The other vessel, called the floodwater storage tank (FWST), is a vertically standing vessel approximately 40 feet high and 10 feet in diameter with a 1/2-inch wall thickness. The FWST support skirt is square (in plan views) with its corners intersecting the ellipsoidal bottom head near the knuckle region

Modeling of laser welding of steel and titanium plates with a composite insert

Science.gov (United States)

Isaev, V. I.; Cherepanov, A. N.; Shapeev, V. P.

2017-10-01

A 3D model of laser welding proposed before by the authors was extended to the case of welding of metallic plates made of dissimilar materials with a composite multilayer intermediate insert. The model simulates heat transfer in the welded plates and takes into account phase transitions. It was proposed to select the composition of several metals and dimensions of the insert to avoid the formation of brittle intermetallic phases in the weld joint negatively affecting its strength properties. The model accounts for key physical phenomena occurring during the complex process of laser welding. It is capable to calculate temperature regimes at each point of the plates. The model can be used to select the welding parameters reducing the risk of formation of intermetallic plates. It can forecast the dimensions and crystalline structure of the solidified melt. Based on the proposed model a numerical algorithm was constructed. Simulations were carried out for the welding of titanium and steel plates with a composite insert comprising four different metals: copper and niobium (intermediate plates) with steel and titanium (outer plates). The insert is produced by explosion welding. Temperature fields and the processes of melting, evaporation, and solidification were studied.

Neutron irradiation effects in pressure vessel steels and weldments

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Ultimate Pressure Capacity of Prestressed Concrete Containment Vessels with Steel Fibers

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-15

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

Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Frey, G.N.

1985-01-01

The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

75 FR 4779 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From Italy: Preliminary Results of...

Science.gov (United States)

2010-01-29

...-Quality Steel Plate Products From Italy: Preliminary Results of Antidumping Duty Administrative Review... administrative review of the antidumping duty order on certain cut-to-length carbon- quality steel plate products... that the Department conduct an administrative review of its sales and entries of subject merchandise...

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

International Nuclear Information System (INIS)

Houze, M.

2002-12-01

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

Development of a surface topography instrument for automotive textured steel plate

Science.gov (United States)

Wang, Zhen; Wang, Shenghuai; Chen, Yurong; Xie, Tiebang

2010-08-01

The surface topography of automotive steel plate is decisive to its stamping, painting and image clarity performances. For measuring this kind of surface topography, an instrument has been developed based on the principle of vertical scanning white light microscopy interference principle. The microscopy interference system of this instrument is designed based on the structure of Linnik interference microscopy. The 1D worktable of Z direction is designed and introduced in details. The work principle of this instrument is analyzed. In measuring process, the interference microscopy is derived as a whole and the measured surface is scanned in vertical direction. The measurement accuracy and validity is verified by templates. Surface topography of textured steel plate is also measured by this instrument.

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.

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

International Nuclear Information System (INIS)

Whitman, G.D.

1975-01-01

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

Apparent embrittlement saturation and radiation mechanisms of reactor pressure vessel steels

International Nuclear Information System (INIS)

Pachur, D.

1981-01-01

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

Internal stresses in steel plate generated by shape memory alloy inserts

International Nuclear Information System (INIS)

Malard, B.; Pilch, J.; Sittner, P.; Davydov, V.; Sedlák, P.; Konstantinidis, K.; Hughes, D.J.

2012-01-01

Graphical abstract: Display Omitted Highlights: ► Thermoresponsive internal stresses introduced into steel by embedding SMA inclusions. ► Neutron strain scanning on steel plate coupons with NiTi inserts at 21 °C and 130 °C. ► Internal stress field in steel evaluated directly from strains and by FE simulation. ► Internal stress generation by SMA insert resistant to thermal and mechanical fatigue. - Abstract: Neutron strain scanning was employed to investigate the internal stress fields in steel plate coupons with embedded prestrained superelastic NiTi shape memory alloy inserts. Strain fields in steel were evaluated at T = 21 °C and 130 °C on virgin coupons as well as on mechanically and thermally fatigued coupons. Internal stress fields were evaluated by direct calculation of principal stress components from the experimentally measured lattice strains as well as by employing an inverse finite element modeling approach. It is shown that if the NiTi inserts are embedded into the elastic steel matrix following a carefully designed technological procedure, the internal stress fields vary with temperature in a reproducible and predictable way. It is estimated that this mechanism of internal stress generation can be safely applied in the temperature range from −20 °C to 150 °C and is relatively resistant to thermal and mechanical fatigue. The predictability and fatigue endurance of the mechanism are of essential importance for the development of future smart metal matrix composites or smart structures with embedded shape memory alloy components.

Applicability of JIS SPV 50 steel to primary containment vessels of nuclear power stations

International Nuclear Information System (INIS)

Iida, K.; Ishikawa, K.; Satoh, M.; Soya, I.

1980-01-01

The fracture toughness of JIS SPV 50 steel and its weldment has been examined in order to verify the applicability of these materials to primary containment vessels of nuclear power stations. Test results were evaluated using elastic plastic fracture mechanics through the COD and the J integral concepts for non ductile fracture initiation characteristics. Linear fracture mechanics was employed for propagation arrest characteristics. Results showed that the materials tested here have a sufficient fracture toughness to prevent nonductile fracture and that this steel is a suitable material for use in construction of primary containment vessels of nuclear power stations. (author)

Preliminary results of steel containment vessel model test

International Nuclear Information System (INIS)

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

1997-01-01

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

Experiment and simulation analysis of roll-bonded Q235 steel plate

International Nuclear Information System (INIS)

Zhao, G.; Huang, Q.; Zhou, C.; Zhang, Z.; Ma, L.; Wang, X.

2016-01-01

Heavy-gauge Q235 steel plate was roll bonded, and the process was simulated using MARC software. Ultrasonic testing results revealed the presence of cracks and lamination defects in an 80-mm clad steel sheet, especially at the head and tail of the steel plate. There were non-uniform ferrite + pearlite microstructures and unbound areas at a bond interface. Through scanning electron microscopy analysis, long cracks and additional inclusions in the cracks were observed at the interface. A fracture analysis revealed non-uniform inclusions that pervaded the interface. Moreover, MARC simulations demonstrated that there was little equivalent strain at the centre of the slab during the first rolling pass. The equivalent centre increased to 0.5 by the fourth rolling pass. Prior to the final pass, the equivalent strain was not consistent across the thickness direction, preventing bonding interfaces from forming consistent deformation and decreasing the residual stress. The initial rolling reduction rate should not be very small (e.g. 5%) as it is averse to the coordination of rolling deformation. Such rolling processes are averse to the rolling bond. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

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

Cytotoxicity difference of 316L stainless steel and titanium reconstruction plate

OpenAIRE

Ni Putu Mira Sumarta; Coen Pramono Danudiningrat; Ester Arijani Rachmat; Pratiwi Soesilawati

2011-01-01

Background: Pure titanium is the most biocompatible material today and used as a gold standard for metallic implants. However, stainless steel is still being used as implants because of its strength, ductility, lower price, corrosion resistant and biocompatibility. Purpose: This study was done to revealed the cytotoxicity difference between reconstruction plate made of 316L stainless steel and of commercially pure (CP) titanium in baby hamster kidney-21 (BHK-21) fibroblast culture through MTT...

Crack propagation on spherical pressure vessels

International Nuclear Information System (INIS)

Lebey, J.; Roche, R.

1975-01-01

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

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

International Nuclear Information System (INIS)

Moranchel y Rodriguez, M.; Garcia B, A.; Longoria G, L. C.

2010-09-01

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)

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

International Nuclear Information System (INIS)

Machatti, H.

1973-01-01

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

Development of stress correction formulae for heat formed steel plates

Directory of Open Access Journals (Sweden)

Hyung Kyun Lim

2018-03-01

Full Text Available The heating process such as line heating, triangular heating and so on is widely used in plate forming of shell plates found in bow and stern area of outer shell in a ship. Local shrinkage during heating process is main physical phenomenon used in plate forming process. As it is well appreciated, the heated plate undergoes the change in material and mechanical properties around heated area due to the harsh thermal process. It is, therefore, important to investigate the changes of physical and mechanical properties due to heating process in order to use them plate the design stage of shell plates. This study is concerned with the development of formula of plastic hardening constitutive equation for steel plate on which line heating is applied. In this study the stress correction formula for the heated plate has been developed based on the numerical simulation of tension test with varying plate thickness and heating speed through the regression analysis of multiple variable case. It has been seen the developed formula shows very good agreement with results of numerical simulation. This paper ends with usefulness of the present formula in examining the structural characteristic of ship's hull. Keywords: Heat input, Heat transfer analysis, Line heating, Shell plate, Stress correction, Thermo-elasto-plastic analysis

The Effect of the Width of an Aluminum Plate on a Bouncing Steel Ball

Directory of Open Access Journals (Sweden)

Christine Hathaway

2013-01-01

Full Text Available The effect of the distance between clamping supports of an aluminum alloy plate on the coefficient of restitution of a bouncing steel ball was investigated. The plate was supported on two wooden blocks with a meter stick secured on either side. A steel ball was dropped from a constant height and a motion detector was used to find the coefficient of restitution. Measurements were made with the wooden blocks at a range of distances. It was found that as the distance between the wooden blocks increased, the coefficient of restitution decreased linearly

Brazing open cell reticulated copper foam to stainless steel tubing with vacuum furnace brazed gold/indium alloy plating

Science.gov (United States)

Howard, Stanley R [Windsor, SC; Korinko, Paul S [Aiken, SC

2008-05-27

A method of fabricating a heat exchanger includes brush electroplating plated layers for a brazing alloy onto a stainless steel tube in thin layers, over a nickel strike having a 1.3 .mu.m thickness. The resultant Au-18 In composition may be applied as a first layer of indium, 1.47 .mu.m thick, and a second layer of gold, 2.54 .mu.m thick. The order of plating helps control brazing erosion. Excessive amounts of brazing material are avoided by controlling the electroplating process. The reticulated copper foam rings are interference fit to the stainless steel tube, and in contact with the plated layers. The copper foam rings, the plated layers for brazing alloy, and the stainless steel tube are heated and cooled in a vacuum furnace at controlled rates, forming a bond of the copper foam rings to the stainless steel tube that improves heat transfer between the tube and the copper foam.

Design of the ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

1995-01-01

The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

Effect of direct quenching on the microstructure and mechanical properties of the lean-chemistry HSLA-100 steel plates

International Nuclear Information System (INIS)

Dhua, S.K.; Sen, S.K.

2011-01-01

Highlights: → Direct-quenched and tempered (DQT) steels gives better mechanical properties. → Fine Cu and Nb (C, N) precipitates enhance matrix strengthening and tempering resistance. → Boron promotes hardenability, but low temperature Charpy impact toughness gets affected. → Mechanical properties equivalent to HSLA-100 steel is achieved by directly quenched leaner chemistry alloys. - Abstract: The influence of direct quenching on structure-property behavior of lean chemistry HSLA-100 steels was studied. Two laboratory heats, one containing Cu and Nb (C:0.052, Mn:0.99, Cu:1.08, Nb:0.043, Cr:0.57, Ni:1.76, Mo:0.55 pct) and the other containing Cu, Nb and B (C:0.04, Mn:1.02, Cu:1.06, Nb:0.036, Cr:0.87, Ni:1.32, Mo:0.41, B:0.002 percent) were hot-rolled into 25 and 12.5 mm thick plates by varying finish-rolling temperatures. The plates were heat-treated by conventional reheat quenching and tempering (RQT), as well as by direct quenching and tempering (DQT) techniques. In general, direct-quench and tempered plates of Nb-Cu heat exhibited good strength (yield strength ∼ 900 MPa) and low-temperature impact toughness (average: 74 J at -85 deg. C); the Charpy V-notch impact energies were marginally lower than conventional HSLA-100 steel. In Nb-Cu-B heat, impact toughness at low-temperature was inferior owing to boron segregation at grain boundaries. Transmission electron microscopy (TEM) and scanning auger microprobe (SAM) analysis confirmed existence of borocarbides at grain boundaries in this steel. In general, for both the steels, the mechanical properties of the direct-quench and tempered plates were found to be superior to reheat quench and tempered plates. A detailed transmission electron microscopy study revealed presence of fine Cu and Nb (C, N) precipitates in these steels. It was also observed that smaller martensite inter-lath spacing, finer grains and precipitates in direct-quench and tempered plates compared to the reheat quench and tempered plates

Numerical simulation of projectile impact on mild steel armour plates using LS-DYNA, Part II: Parametric studies

OpenAIRE

Raguraman, M; Deb, A; Gupta, NK; Kharat, DK

2008-01-01

In Part I of the current two-part series, a comprehensive simulation-based study of impact of Jacketed projectiles on mild steel armour plates has been presented. Using the modelling procedures developed in Part I, a number of parametric studies have been carried out for the same mild steel plates considered in Part I and reported here in Part II. The current investigation includes determination of ballistic limits of a given target plate for different projectile diameters and impact velociti...

Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

International Nuclear Information System (INIS)

Ghoneim, M.M.

1987-05-01

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

Underwater Shock Response of Circular HSLA Steel Plates

Directory of Open Access Journals (Sweden)

R. Rajendran

2000-01-01

Full Text Available Studies on shock response of circular plates subjected to underwater explosion is of interest to ship designers. Non-contact underwater explosion experiments were carried out on air backed circular High Strength Low Alloy (HSLA steel plates of 4 mm thickness and 290 mm diameter. The experiments were carried out in two phases. In the first phase, strain gauges were fixed at intervals of 30 mm from the centre of the plate and strains were recorded for the shock intensity gradually increasing to yielding. Semi-analytical models were derived for the elastic strain prediction which showed good agreement with the experiments. Dynamic yield stress and the shock factor for yielding were established. In the second phase, individual plates were subjected to increasing shock severity until fracture and the apex bulge depth and the thickness strains were measured. Empirical models were derived to predict the plastic deformation which were validated through a fresh set of experiments. Analysis of the fractured surface by visual examination showed that there was slant fracture indicating ductile mode of failure and the same was corroborated by Scanning Electron Microscopic (SEM examination.

Analysis of the Behaviour of Semi Rigid Steel End Plate Connections

Directory of Open Access Journals (Sweden)

Bahaz A.

2018-01-01

Full Text Available The analysis of steel-framed building structures with full strength beam to column joints is quite standard nowadays. Buildings utilizing such framing systems are widely used in design practice. However, there is a growing recognition of significant benefits in designing joints as partial strength/semi-rigid. The design of joints within this partial strength/semi-rigid approach is becoming more and more popular. This requires the knowledge of the full nonlinear moment-rotation behaviour of the joint, which is also a design parameter. The rotational behaviour of steel semi rigid connections can be studied using the finite element method for the following three reasons: i such models are inexpensive; ii they allow the understanding of local effects, which are difficult to measure accurately physically, and iii they can be used to generate extensive parametric studies. This paper presents a three-dimensional finite element model using ABAQUS software in order to identify the effect of different parameters on the behaviour of semi rigid steel beam to column end plate connections. Contact and sliding between different elements, bolt pretension and geometric and material non-linearity are included in this model. A parametric study is conducted using a model of two end-plate configurations: flush and extended end plates. The studied parameters were as follows: bolts type, end plate thickness and column web stiffener. Then, the model was calibrated and validated with experimental results taken from the literature and with the model proposed by Eurocode3. The procedure for determining the moment–rotation curve using finite element analysis is also given together with a brief explanation of how the design moment resistance and the initial rotational stiffness of the joint are obtained.

Full title: Biomechanical comparison between stainless steel, titanium and carbon-fiber reinforced polyetheretherketone volar locking plates for distal radius fractures.

Science.gov (United States)

Mugnai, Raffaele; Tarallo, Luigi; Capra, Francesco; Catani, Fabio

2018-05-25

As the popularity of volar locked plate fixation for distal radius fractures has increased, so have the number and variety of implants, including variations in plate design, the size and angle of the screws, the locking screw mechanism, and the material of the plates. carbon-fiber reinforced polyetheretherketone (CFR-PEEK) plate features similar biomechanical properties to metallic plates, representing, therefore, an optimal alternative for the treatment of distal radius fractures. three different materials-composed plates were evaluated: stainless steel volar lateral column (Zimmer); titanium DVR (Hand Innovations); CFR-PEEK DiPHOS-RM (Lima Corporate). Six plates for each type were implanted in sawbones and an extra-articular rectangular osteotomy was created. Three plates for each material were tested for load to failure and bending stiffness in axial compression. Moreover, 3 constructs for each plate were evaluated after dynamically loading for 6000 cycles of fatigue. the mean bending stiffness pre-fatigue was significantly higher for the stainless steel plate. The titanium plate yielded the higher load to failure both pre and post fatigue. After cyclic loading, the bending stiffness increased by a mean of 24% for the stainless steel plate; 33% for the titanium; and 17% for the CFR-PEEK plate. The mean load to failure post-fatigue increased by a mean of 10% for the stainless steel and 14% for CFR-PEEK plates, whereas it decreased (-16%) for the titanium plate. Statistical analysis between groups reported significant values (p plastic deformation, and lower load to failure. N/A. Copyright © 2018. Published by Elsevier Masson SAS.

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

International Nuclear Information System (INIS)

Pareige, P.

1996-04-01

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

Wetting Behavior of Molten AZ61 Magnesium Alloy on Two Different Steel Plates Under the Cold Metal Transfer Condition

Directory of Open Access Journals (Sweden)

ZENG Cheng-zong

2017-04-01

Full Text Available The wetting behavior and interfacial microstructures of molten magnesium AZ61 alloy on the surface of two different Q235 and galvanized steel plates under the condition of cold metal transfer were investigated by using dynamic sessile drop method. The results show that the wetting behavior is closely related to the wire feed speed. Al-Fe intermetallic layer was observed whether the substrate is Q235 steel or galvanized steel, and the formation of Al-Fe intermetallic layer should satisfy the thermodynamic condition of such Mg-Al/Fe system. The wettability of molten AZ61 magnesium alloy is improved with the increase of wire feed speed whether on Q235 steel surface or on galvanized steel surface, good wettability on Q235 steel surface is due to severe interface reaction when wire feed speed increases, good wettability on galvanized steel surface is attributed to the aggravating zinc volatilization. When the wire feed speed is ≤10.5m·min-1, the wettability of Mg alloy on Q235 steel plate is better than on galvanized steel plate. However, Zn vapor will result in instability for metal transfer process.

Effects of thermal ageing on toughness properties of pressure vessel steel

International Nuclear Information System (INIS)

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

2015-01-01

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

75 FR 61699 - Stainless Steel Plate in Coils From Belgium, Italy, South Africa, South Korea, and Taiwan: Final...

Science.gov (United States)

2010-10-06

...-831, and A-583-830] Stainless Steel Plate in Coils From Belgium, Italy, South Africa, South Korea, and... steel plate in coils (SSPC) from Belgium, Italy, South Africa, South Korea, and Taiwan, pursuant to... sunset reviews of the antidumping duty orders on SSPC from Belgium, Italy, South Africa, South Korea, and...

Experimental study of the effect of neutron radiation on pressurised water reactor vessel steel resilience - First part

International Nuclear Information System (INIS)

Verdeau, Jean-Jacques

1969-12-01

After having outlined the importance of the embrittlement of vessel steels by neutrons during the exploitation of pressurised water reactors, the author reports a set of tests which aimed at determining the effect of neutron irradiation on vessel steel resilience for operated, under construction or projected pressurized water reactors. He also tries to highlight the influence of irradiation temperature and of initial thermal treatments, and to look for a restoration thermal treatment of neutron-induced damages which could be applied to the considered vessels. Tests were performed on V Charpy resilience samples. Some samples have been irradiated by the Pile Department of the Grenoble CEN and then broken by the Laboratory of very high activity, whereas other samples have been irradiated in a prototype vessel and broken by a Cadarache department. The author presents characteristics of the studied steels (chemical compositions, thermal treatments), describes sample irradiation conditions, and the method of assessment of the transition temperature after irradiation, presents experimental results, discusses their interpretation, and presents future tests to be performed [fr

Safety of steel vessel Magnox pressure circuits

International Nuclear Information System (INIS)

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

1991-01-01

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

X-Ray diffraction technique applied to study of residual stresses after welding of duplex stainless steel plates

International Nuclear Information System (INIS)

Monin, Vladimir Ivanovitch; Assis, Joaquim Teixeira de; Lopes, Ricardo Tadeu; Turibus, Sergio Noleto; Payao Filho, Joao C.

2014-01-01

Duplex stainless steel is an example of composite material with approximately equal amounts of austenite and ferrite phases. Difference of physical and mechanical properties of component is additional factor that contributes appearance of residual stresses after welding of duplex steel plates. Measurements of stress distributions in weld region were made by X-ray diffraction method both in ferrite and austenite phases. Duplex Steel plates were joined by GTAW (Gas Tungsten Arc Welding) technology. There were studied longitudinal and transverse stress components in welded butt joint, in heat affected zone (HAZ) and in points of base metal 10 mm from the weld. Residual stresses measured in duplex steel plates jointed by welding are caused by temperature gradients between weld zone and base metal and by difference of thermal expansion coefficients of ferrite and austenite phases. Proposed analytical model allows evaluating of residual stress distribution over the cross section in the weld region. (author)

Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

Lopez Jimenez, J.

1978-01-01

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

Fatigue in Welded High-Strength Steel Plate Elements under Stochastic Loading

DEFF Research Database (Denmark)

Agerskov, Henning; Petersen, R.I.; Martinez, L. Lopez

1999-01-01

The present project is a part of an investigation on fatigue in offshore structures in high-strength steel. The fatigue life of plate elements with welded attachments is studied. The material used has a yield stress of ~ 810-840 MPa, and high weldability and toughness properties. Fatigue test...... series with constant amplitude loading and with various types of stochastic loading have been carried through on test specimens in high-strength steel, and - for a comparison - on test specimens in conventional offshore structural steel with a yield stress of ~ 400-410 MPa.A comparison between constant...... amplitude and variable amplitude fatigue test results shows shorter fatigue lives in variable amplitude loading than should be expected from the linear fatigue damage accumulation formula. Furthermore, in general longer fatigue lives were obtained for the test specimens in high-strength steel than those...

Destructive examination of test plates 1 and 2 of the defects detection trials

International Nuclear Information System (INIS)

Crutzen, S.; Buergers, W.; Violin, F.; Di Piazza, L.; Cowburn, K.; Sargent, T.

1983-01-01

A further phase of the UKAEA defect detection trials (described previously) with PWR pressure vessel steels is reported. The evaluation of NDT exercise results must be based on destructive examination of the plates used during the exercise. Tests are described and results given. (U.K.)

Characteristics of martensite as a function of the Ms temperature in low-carbon armour steel plates

International Nuclear Information System (INIS)

Maweja, Kasonde; Stumpf, Waldo; Berg, Nic van der

2009-01-01

The microstructure, morphology, crystal structure and surface relief of martensite in a number of experimental armour steel plates with different M s temperatures were analysed. Atomic force microscopy, thin foil transmission electron microscopy and scanning electron microscopy allowed the identification of three groups of low-carbon martensitic armour steels. The investigation showed that the size of individual martensite products (plates or packets, laths or blocks) increases as the M s temperature increases. Comparison of ballistic performances suggests that the morphology (plate or lath) and size of the individual martensite products dictate the effective 'grain size' in resisting fracture or perforation due to ballistic impact.

Fire resistance of a steel plate reinforced concrete bearing wall

International Nuclear Information System (INIS)

Kodaira, Akio; Kanchi, Masaki; Fujinaka, Hideo; Akita, Shodo; Ozaki, Masahiko

2003-01-01

Samples from a steel plate reinforced concrete bearing wall composed of concrete slab sandwiched between studded steel plates, were subjected to loaded fire resistance tests. There were two types of specimens: some were 1800 mm high while the rest were 3000 mm high ; thickness and width were the same for all specimens, at 200 mm and 800 mm, respectively. Under constant load conditions, one side of each specimen was heated along the standard fire-temperature curve. The results enabled us to approximate the relationship between the ratio of working load to concrete strength N/(Ac x c σ b) and the fire resistance time (t: minutes), as equation (1) for the 1800 mm - high specimen, and equation (2) for the 3000 mm - high specimen. N/(Ac x c σ b) = 2.21 x (1/t) 0.323 (1), .N/(Ac x c σ b) 2.30 x (1/t) 0.378 (2) In addition, the temperature of the unheated side of the specimens was 100degC at 240 minutes of continuous heating, clearly indicating that there was sufficient heat insulation. (author)

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

International Nuclear Information System (INIS)

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

1990-11-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

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

International Nuclear Information System (INIS)

Zhang, K.D.; Wang, W.

1987-01-01

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

Vacuum vessel for thermonuclear device

International Nuclear Information System (INIS)

Kikuchi, Mitsuru; Nagashima, Keisuke; Suzuki, Masaru; Onozuka, Masaki.

1997-01-01

A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

Vacuum vessel for thermonuclear device

Energy Technology Data Exchange (ETDEWEB)

Kikuchi, Mitsuru; Nagashima, Keisuke [Japan Atomic Energy Research Inst., Tokyo (Japan); Suzuki, Masaru; Onozuka, Masaki

1997-07-11

A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

Containment liner plate anchors and steel embedments test results

International Nuclear Information System (INIS)

Chang-Lo, P.L.; Johnson, T.E.; Pfeifer, B.W.

1977-01-01

This paper summarizes test data on shear load and deformation capabilities for liner plate line anchors and structural steel embedments in reinforced and prestressed concrete nuclear containments. Reinforced and prestressed nuclear containments designed and constructed in the United States are lined with a minimum of 0.64 cm steel plate. The liner plates are anchored by the use of either studs or structural members (line anchors) which usually run in the vertical direction. This paper will only address line anchors. Static load versus displacement test data is necessary to assure that the design is adequate for the maximum loads. The test program for the liner anchors had the following major objectives: determine load versus displacement data for a variety of anchors considering structural tees and small beams with different weld configurations, from the preceding tests, determine which anchors would lead to an economical and extremely safe design and test these anchors for cyclic loads resulting from thermal fluctuations. Various concrete embeds in the containment and other structures are subjected to loads such as pipe rupture which results in shear. Since many of the loads are transient by nature, it is necessary to know the load-displacement relationship so that the energy absorption can be determined. The test program for the embeds had the following objectives: determine load-displacement relationship for various size anchors from 6.5 cm 2 to 26 cm 2 with maximum capacities of approximately 650 kN; determine the effect of various anchor width-to-thickness ratios for the same shear area

75 FR 10207 - Certain Cut-to-Length Carbon-Quality Steel Plate From the Republic of Korea: Final Results of...

Science.gov (United States)

2010-03-05

...-Quality Steel Plate From the Republic of Korea: Final Results of Antidumping Duty Administrative Review... cut-to-length carbon-quality steel plate from the Republic of Korea and the intent to rescind the... 1, 2008, through January 31, 2009. We have rescinded the review with respect to one company and we...

Simulation analysis of impact tests of steel plate reinforced concrete and reinforced concrete slabs against aircraft impact and its validation with experimental results

International Nuclear Information System (INIS)

Sadiq, Muhammad; Xiu Yun, Zhu; Rong, Pan

2014-01-01

Highlights: • Simulation analysis is carried out with two constitutive concrete models. • Winfrith model can better simulate nonlinear response of concrete than CSCM model. • Performance of steel plate concrete is better than reinforced concrete. • Thickness of safety related structures can be reduced by adopting steel plates. • Analysis results, mainly concrete material models should be validated. - Abstract: The steel plate reinforced concrete and reinforced concrete structures are used in nuclear power plants for protection against impact of an aircraft. In order to compare the impact resistance performance of steel plate reinforced concrete and reinforced concrete slabs panels, simulation analysis of 1/7.5 scale model impact tests is carried out by using finite element code ANSYS/LS-DYNA. The damage modes of all finite element models, velocity time history curves of the aircraft engine and damage to aircraft model are compared with the impact test results of steel plate reinforced concrete and reinforced concrete slab panels. The results indicate that finite element simulation results correlate well with the experimental results especially for constitutive winfrith concrete model. Also, the impact resistance performance of steel plate reinforced concrete slab panels is better than reinforced concrete slab panels, particularly the rear face steel plate is very effective in preventing the perforation and scabbing of concrete than conventional reinforced concrete structures. In this way, the thickness of steel plate reinforced concrete structures can be reduced in important structures like nuclear power plants against impact of aircraft. It also demonstrates the methodology to validate the analysis procedure with experimental and analytical studies. It may be effectively employed to predict the precise response of safety related structures against aircraft impact

Casting of Hearth Plates from High-chromium Steel

Directory of Open Access Journals (Sweden)

Drotlew A.

2014-12-01

Full Text Available The paper presents the results of studies on the development of manufacturing technologies to cast hearth plates operating in chamber furnaces for heat treatment. Castings made from the heat-resistant G-X40CrNiSi27-4 steel were poured in hand-made green sand molds. The following operations were performed: computer simulation to predict the distribution of internal defects in castings produced by the above mentioned technology with risers bare and coated with exothermic and insulating sleeves, analysis of each variant of the technology, and manufacture of experimental castings. As a result of the conducted studies and analysis it was found that the use of risers with exothermic sleeves does not affect to a significant degree the quality of the produced castings of hearth plates, but it significantly improves the metal yield.

Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

Energy Technology Data Exchange (ETDEWEB)

Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

Structural design of nuclear power plant using stiffened steel plate concrete structure

International Nuclear Information System (INIS)

Moon, Ilhwan; Kim, Sungmin; Mun, Taeyoup; Kim, Keunkyeong; Sun, Wonsang

2009-01-01

Nuclear power is an alternative energy source that is conducive to mitigate the environmental strains. The countries having nuclear power plants are encouraging research and development sector to find ways to construct safer and more economically feasible nuclear power plants. Modularization using Steel Plate Concrete(SC) structure has been proposed as a solution to these efforts. A study of structural modules using SC structure has been performed for shortening of construction period and enhancement of structural safety of NPP structures in Korea. As a result of the research, the design code and design techniques based on limit state design method has been developed. The design code has been developed through various structural tests and theoretical studies, and it has been modified by application design of SC structure for NPP buildings. The code consists of unstiffened SC wall design, stiffened SC wall design, Half-SC slab design, stud design, connection design and so on. The stiffened steel plate concrete(SSC) wall is SC structure whose steel plates with ribs are composed on both sides of the concrete wall, and this structure was developed for improved constructability and safety of SC structure. This paper explains a design application of SC structure for a sample building specially devised to reflect all of major structural properties of main buildings of APR1400. In addition, Stiffening effect of SSC structure is evaluated and structural efficiency of SSC structure is verified in comparison with that of unstiffened SC structure. (author)

Study of cladding toughness in a pressure vessel steel water reactor

International Nuclear Information System (INIS)

Soulat, P.; Al Mundheri, M.

1984-12-01

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

Steel Plate Shear Walls: Efficient Structural Solution for Slender High-Rise in China

International Nuclear Information System (INIS)

Mathias, Neville; Long, Eric; Sarkisian, Mark; Huang Zhihui

2008-01-01

The 329.6 meter tall 74-story Jinta Tower in Tianjin, China, is expected, when complete, to be the tallest building in the world with slender steel plate shear walls used as the primary lateral load resisting system. The tower has an overall aspect ratio close to 1:8, and the main design challenge was to develop an efficient lateral system capable of resisting significant wind and seismic lateral loads, while simultaneously keeping wind induced oscillations under acceptable perception limits. This paper describes the process of selection of steel plate shear walls as the structural system, and presents the design philosophy, criteria and procedures that were arrived at by integrating the relevant requirements and recommendations of US and Chinese codes and standards, and current on-going research

Neutron irradiation embrittlement of reactor pressure vessel steels

International Nuclear Information System (INIS)

Steele, L.E.

1975-01-01

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

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

International Nuclear Information System (INIS)

Klaar, H.J.

1975-01-01

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

Corrosion Characterization in Nickel Plated 110 ksi Low Alloy Steel and Incoloy 925: An Experimental Case Study

Science.gov (United States)

Thomas, Kiran; Vincent, S.; Barbadikar, Dipika; Kumar, Shresh; Anwar, Rebin; Fernandes, Nevil

2018-04-01

Incoloy 925 is an age hardenable Nickel-Iron-Chromium alloy with the addition of Molybdenum, Copper, Titanium and Aluminium used in many applications in oil and gas industry. Nickel alloys are preferred mostly in corrosive environments where there is high concentration of H2S, CO2, chlorides and free Sulphur as sufficient nickel content provides protection against chloride-ion stress-corrosion cracking. But unfortunately, Nickel alloys are very expensive. Plating an alloy steel part with nickel would cost much lesser than a part make of nickel alloy for large quantities. A brief study will be carried out to compare the performance of nickel plated alloy steel with that of an Incoloy 925 part by conducting corrosion tests. Tests will be carried out using different coating thicknesses of Nickel on low alloy steel in 0.1 M NaCl solution and results will be verified. From the test results we can confirm that Nickel plated low alloy steel is found to exhibit fairly good corrosion in comparison with Incoloy 925 and thus can be an excellent candidate to replace Incoloy materials.

Study on the application of 50 mm thick welded joints without PWHT for containment vessels

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro (Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan)); Iida, Kunihiro (Shibaura Inst. of Tech., Dept. of Mechanical Engineering, Tokyo (Japan)); Satou, Masanobu (Mitsubishi Heavy Industries. Ltd., Tkasago Research and Development Center (Japan))

1992-01-01

In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.).

Study on the application of 50 mm thick welded joints without PWHT for containment vessels

International Nuclear Information System (INIS)

Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro; Iida, Kunihiro; Satou, Masanobu

1992-01-01

In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.)

Energy-Dissipation Performance of Combined Low Yield Point Steel Plate Damper Based on Topology Optimization and Its Application in Structural Control

Directory of Open Access Journals (Sweden)

Haoxiang He

2016-01-01

Full Text Available In view of the disadvantages such as higher yield stress and inadequate adjustability, a combined low yield point steel plate damper involving low yield point steel plates and common steel plates is proposed. Three types of combined plate dampers with new hollow shapes are proposed, and the specific forms include interior hollow, boundary hollow, and ellipse hollow. The “maximum stiffness” and “full stress state” are used as the optimization objectives, and the topology optimization of different hollow forms by alternating optimization method is to obtain the optimal shape. Various combined steel plate dampers are calculated by finite element simulation, the results indicate that the initial stiffness of the boundary optimized damper and interior optimized damper is lager, the hysteresis curves are full, and there is no stress concentration. These two types of optimization models made in different materials rations are studied by numerical simulation, and the adjustability of yield stress of these combined dampers is verified. The nonlinear dynamic responses, seismic capacity, and damping effect of steel frame structures with different combined dampers are analyzed. The results show that the boundary optimized damper has better energy-dissipation capacity and is suitable for engineering application.

Degradation of the compressive strength of unstiffened/stiffened steel plates due to both-sides randomly distributed corrosion wastage

Directory of Open Access Journals (Sweden)

Zorareh Hadj Mohammad

Full Text Available The paper addresses the problem of the influence of randomly distributed corrosion wastage on the collapse strength and behaviour of unstiffened/stiffened steel plates in longitudinal compression. A series of elastic-plastic large deflection finite element analyses is performed on both-sides randomly corroded steel plates and stiffened plates. The effects of general corrosion are introduced into the finite element models using a novel random thickness surface model. Buckling strength, post-buckling behaviour, ultimate strength and post-ultimate behaviour of the models are investigated as results of both-sides random corrosion.

Irradiated dynamic fracture toughness of ASTM A533, Grade B, Class 1 steel plate and submerged arc weldment. Heavy section steel technology program technical report No. 41

International Nuclear Information System (INIS)

Davidson, J.A.; Ceschini, L.J.; Shogan, R.P.; Rao, G.V.

1976-10-01

As a result of the Heavy Section Steel Technology Program (HSST), sponsored by the Nuclear Regulatory Commission, Westinghouse Electric Corporation conducted dynamic fracture toughness tests on irradiated HSST Plate 02 and submerged arc weldment material. Testing performed at the Westinghouse Research and Development Laboratory in Pittsburgh, Pennsylvania, included 0.394T compact tension, 1.9T compact tension, and 4T compact tension specimens. This data showed that, in the transition region, dynamic test procedures resulted in lower (compared to static) fracture toughness results, and that weak direction (WR) oriented specimen data were lower than the strong direction (RW) oriented specimen results. Irradiated lower-bound fracture toughness results of the HSST Program material were well above the adjusted ASME Section III K/sub IR/ curve. An irradiated and nonirradiated 4T-CT specimen was tested during a fracture toughness test as a preliminary study to determine the effect of irradiation on the acoustic emission-stress intensity factor relation in pressure vessel grade steel. The results indicated higher levels of acoustic emission activity from the irradiated sample as compared to the unirradiated one at a given stress intensity factor (K) level

Interaction between molten corium UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} and VVER vessel steel

Energy Technology Data Exchange (ETDEWEB)

Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A. [Alexandrov Sci Res Technol Inst, Sosnovyi Bor (Russian Federation); Gusarov, V. V.; Almiashev, V. I. [Russian Acad Sci, Inst Silicate Chem, St Petersburg (Russian Federation); Lopukh, D. B. [SPb State Electrotech Univ LETI SPbGETU, St Petersburg (Russian Federation); Bottomley, D. [Joint Res Ctr, Inst Transurane, Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA Saclay, DEN, DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Dresden Rossendorf, Dresden (Germany); Fichot, F. [CEA Cadarache, SEMCA, DPAM, IRSN, St Paul Les Durance (France); Kymalainen, O. [FORTUM Nucl Serv Ltd, Espoo (Finland)

2010-07-01

In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

International Nuclear Information System (INIS)

2010-01-01

The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

Time-of-flight neutron Bragg-edge transmission imaging of microstructures in bent steel plates

Energy Technology Data Exchange (ETDEWEB)

Su, Yuhua, E-mail: yuhua.su@j-parc.jp [J-PARC Center, Japan Atomic Energy Agency, 2-4 Shirakata, Tokai, Ibaraki 319-1195 (Japan); Oikawa, Kenichi; Harjo, Stefanus; Shinohara, Takenao; Kai, Tetsuya; Harada, Masahide; Hiroi, Kosuke [J-PARC Center, Japan Atomic Energy Agency, 2-4 Shirakata, Tokai, Ibaraki 319-1195 (Japan); Zhang, Shuoyuan; Parker, Joseph Don [Neutron R& D Division, CROSS-Tokai, 162-1 Shirakata, Tokai, Ibaraki 319-1106 (Japan); Sato, Hirotaka [Faculty of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Shiota, Yoshinori; Kiyanagi, Yoshiaki [Graduate School of Engineering, Nagoya University, Nagoya, Aichi 464-8603 (Japan); Tomota, Yo [Research Center for Strategic Materials, National Institute for Materials Science, Tsukuba 305-0047 (Japan)

2016-10-15

Neutron Bragg-edge transmission imaging makes it possible to quantitatively visualize the two-dimensional distribution of microstructure within a sample. In order to examine its application to engineering products, time-of-flight Bragg-edge transmission imaging experiments using a pulsed neutron source were performed for plastically bent plates composed of a ferritic steel and a duplex stainless steel. The non-homogeneous microstructure distributions, such as texture, crystalline size, phase volume fraction and residual elastic strain, were evaluated for the cross sections of the bent plates. The obtained results were compared with those by neutron diffraction and electron back scatter diffraction, showing that the Bragg-edge transmission imaging is powerful for engineering use.

Experimental and numerical study of water-filled vessel impacted by flat projectiles

Science.gov (United States)

Zhang, Wei; Ren, Peng; Huang, Wei; Gao, Yu Bo

2014-05-01

To understand the failure modes and impact resistance of double-layer plates separated by water, a flat-nosed projectile was accelerated by a two-stage light gas gun against a water-filled vessel which was placed in an air-filled tank. Targets consisted of a tank made of two flat 5A06 aluminum alloy plates held by a high strength steel frame. The penetration process was recorded by a digital high-speed camera. The same projectile-target system was also used to fire the targets placed directly in air for comparison. Parallel numerical tests were also carried out. The result indicated that experimental and numerical results were in good agreement. Numerical simulations were able to capture the main physical behavior. It was also found that the impact resistance of double layer plates separated by water was lager than that of the target plates in air. Tearing was the main failure models of the water-filled vessel targets which was different from that of the target plates in air where the shear plugging was in dominate.

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-07-01

A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

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.

Multilayer graphene for long-term corrosion protection of stainless steel bipolar plates for polymer electrolyte membrane fuel cell

DEFF Research Database (Denmark)

Stoot, Adam Carsten; Camilli, Luca; Spiegelhauer, Susie Ann

2015-01-01

Abstract Motivated by similar investigations recently published (Pu et al., 2015), we report a comparative corrosion study of three sets of samples relevant as bipolar plates for polymer electrolyte fuel cells: stainless steel, stainless steel with a nickel seed layer (Ni/SS) and stainless steel...

Development of ultrasonic testing technique with a large transducer to inspect the containment vessel plates embedded in concrete for corrosion on nuclear power plant (2)

International Nuclear Information System (INIS)

Ishida, Hitoshi

2005-01-01

The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. The purpose of this study is establishment of ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely from the accessible point. Experiments to detect artificial hollows simulating corrosion and stud bolts which hold the mold of concrete on a surface of a carbon steel plate mock-up covered with concrete were carried out with newly made low frequency (0.3MHz and 0.5MHz) 90 degrees refraction angle shear horizontal (SH) wave transducers combined with three active elements, which were equivalent to a 120 mm width element. As the results: (1) The echoes from the artificial hollows with a depth of 19 mm and 9.5mm at a distance of 1.5 m and the stud bolts with a diameter of 8mm at a distance of 0.7 - 1.7m could be discriminated clearly. (2) The multiple echoes bouncing three times between the front side and the back side of the plate, which was equivalent to a distance of about 12m, could be discriminated. (3) A divergence angle and a -6dB divergence angle of the large element (combined three elements) transducer were about 7 degrees and about 3 degrees. (4) The echoes from the hollows with a depth of 9.5m could be detected at a distance of 3.6 m with a reflection at the side wall of the mock-up. (5) It was estimated that the maximum distance of detection of the echo from the stud bolt with a diameter of 8mm was about 2.9 ∼ 3.6 m. Therefore we evaluate that the large element transducer can propagate the SH wave to about a half of a distance to the bottom of the embedded containment vessel and it is possible to detect the defects such as corrosion to a distance of 3.6 m. (author)

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.

Electron beam cladding of titanium on stainless steel plate

International Nuclear Information System (INIS)

Tomie, Michio; Abe, Nobuyuki; Yamada, Masanori; Noguchi, Shuichi.

1990-01-01

Fundamental characteristics of electron beam cladding was investigated. Titanium foil of 0.2mm thickness was cladded on stainless steel plate of 3mm thickness by scanning electron beam. Surface roughness and cladded layer were analyzed by surface roughness tester, microscope, scanning electron microscope and electron probe micro analyzer. Electron beam conditions were discussed for these fundamental characteristics. It is found that the energy density of the electron beam is one of the most important factor for cladding. (author)

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.

The effect of different rutile electrodes on mechanical properties of underwater wet welded AH-36 steel plates

Science.gov (United States)

Winarto, Winarto; Purnama, Dewin; Churniawan, Iwan

2018-04-01

Underwater welding is an important role in the rescue of ships and underwater structures, in case of emergency. In this study, the marine steel plates used are AH-36 steel as parent material. This type of steel is included in the High Strength Low Alloy (HSLA). Electrodes used for welding AH-36 steel plates are commonly the E6013 and E 7024 which are the type of based rutile electrodes. Those electrodes are widely available on the market and they would be compared with the original electrode for underwater which is the type of E7014 with the trade name of Broco UW-CS-1. Welding method used is Shielding Metal Arc Welding (SMAW) with the variation of 5 m and 10 m underwater depth and also varied with the electric current of 120A, 140A and 250A. It was found that hardness value of increased in the area of weld metal and HAZ. HAZ also tends to have the highest hardness compared to both of weld metal and base metal. Non destructive test by radiographed test (RT) on welds showed that there are found welding defects in the form of incomplete penetration on all variations of welding parameters, but there is no porosity defect detected. The results of the hardness tests of underwater wet welded steel plates show that the hardness value of both rutile electrodes (E6013 and E 7024) is apparently similar hardness value compared with the existing commercial electrode (E7014 of Broco UW-CS- 1). The tensile test results of underwater wet welded steel plates show that the use of rutile electrode of E6013 gives a better tensile properties than other rutile electrodes.

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

International Nuclear Information System (INIS)

1998-10-01

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

Measurement of the yield and tensile strengths of neutron-irradiated and post-irradiation recovered vessel steels with notched specimens

International Nuclear Information System (INIS)

Valiente, A.

1996-01-01

Tensile circumferentially notched bars are examined as test specimens for measuring the yield and tensile strengths of nuclear pressure vessel steels under several conditions of irradiation and temperature that a vessel can experience during its service life, including recovery post-irradiation treatment. For all the vessel steels, notch geometries and conditions explored, it has been found that notched specimens fail by plastic collapse, and simple formulae have been derived that allow the yield and tensile strengths to be determined from the yielding and plastic collapse load of a notched specimen. Values measured in this way show good agreement with those measured by the standard tensile test method. (orig.)

Trial manufacturing of titanium-carbon steel composite overpack

International Nuclear Information System (INIS)

Honma, Nobuyuki; Chiba, Takahiko; Tanai, Kenji

1999-11-01

This paper reports the results of design analysis and trial manufacturing of full-scale titanium-carbon steel composite overpacks. The overpack is one of the key components of the engineered barrier system, hence, it is necessary to confirm the applicability of current technique in their manufacture. The required thickness was calculated according to mechanical resistance analysis, based on models used in current nuclear facilities. The Adequacy of the calculated dimensions was confirmed by finite-element methods. To investigate the necessity of a radiation shielding function of the overpack, the irradiation from vitrified waste has been calculated. As a result, it was shown that shielding on handling and transport equipment is a more reasonable and practical approach than to increase thickness of overpack to attain a self-shielding capability. After the above investigation, trial manufacturing of full-scale model of titanium-carbon steel composite overpack has been carried out. For corrosion-resistant material, ASTM Grade-2 titanium was selected. The titanium layer was bonded individually to a cylindrical shell and fiat cover plates (top and bottom) made of carbon steel. For the cylindrical shell portion, a cylindrically formed titanium layer was fitted to the inner carbon steel vessel by shrinkage. For the flat cover plates (top and bottom), titanium plate material was coated by explosive bonding. Electron beam welding and gas metal arc welding were combined to weld of the cover plates to the body. No significant failure was evident from inspections of the fabrication process, and the applicability of current technology for manufacturing titanium-carbon steel composite overpack was confirmed. Future research and development items regarding titanium-carbon steel composite overpacks are also discussed. (author)

Tensile properties of irradiated and fatigue exposed stainless steel DIN X 6 CrNi 1811 (similar to AISI type 304) plate and welded joints

International Nuclear Information System (INIS)

Vries, M.I. de; Schaaf, B. van der; Elen, J.D.

1979-10-01

Test specimens of plate metal and welded joints of stainless steel DIN 1.4948, which is similar to AISI type 304, have been irradiated at 723 K and 823 K up to fluences of 1.10 23 n.m -2 and 5.10 24 n.m -2 (E > 0.1 MeV). These are representative conditions for the SNR-300 reactor vessel and inner components after 16 years of operation. High-rate (depsilon/dt = 1 s -1 ) tensile tests were performed after fatigue exposure up to various fractions of fatigue life (D) ranging from 5% to 95% at the same temperatures as the nominal temperatures of the irradiation series

Experimental Study on Welded Headed Studs Used In Steel Plate-Concrete Composite Structures Compared with Contactless Method of Measuring Displacement

Science.gov (United States)

Kisała, Dawid; Tekieli, Marcin

2017-10-01

Steel plate-concrete composite structures are a new innovative design concept in which a thin steel plate is attached to the reinforced concrete beam by means of welded headed studs. The comparison between experimental studies and theoretical analysis of this type of structures shows that their behaviour is dependent on the load-slip relationship of the shear connectors used to ensure sufficient bond between the concrete and steel parts of the structure. The aim of this paper is to describe an experimental study on headed studs used in steel plate-concrete composite structures. Push-out tests were carried out to investigate the behaviour of shear connectors. The test specimens were prepared according to standard push-out tests, however, instead of I-beam, a steel plate 16 mm thick was used to better reflect the conditions in the real structure. The test specimens were produced in two batches using concrete with significantly different compressive strength. The experimental study was carried out on twelve specimens. Besides the traditional measurements based on LVDT sensors, optical measurements based on the digital image correlation method (DIC) and pattern tracking methods were used. DIC is a full-field contactless optical method for measuring displacements in experimental testing, based on the correlation of the digital images taken during test execution. With respect to conventional methods, optical measurements offer a wider scope of results and can give more information about the material or construction behaviour during the test. The ultimate load capacity and load-slip curves obtained from the experiments were compared with the values calculated based on Eurocodes, American and Chinese design specifications. It was observed that the use of the relationships developed for the traditional steel-concrete composite structures is justified in the case of ultimate load capacity of shear connectors in steel plate-concrete composite structures.

Review on Electroless Plating Ni-P Coatings for Improving Surface Performance of Steel

Science.gov (United States)

Zhang, Hongyan; Zou, Jiaojuan; Lin, Naiming; Tang, Bin

2014-04-01

Electroless plating has been considered as an effective approach to provide protection and enhancement for metallic materials with many excellent properties in engineering field. This paper begins with a brief introduction of the fundamental aspects underlying the technological principles and conventional process of electroless nickel-phosphorus (Ni-P) coatings. Then this paper discusses different electroless nickel plating, including binary plating, ternary composite plating and nickel plating with nanoparticles and rare earth, with the intention of improving the surface performance on steel substrate in recent years in detail. Based on different coating process, the varied features depending on the processing parameters are highlighted. Separately, diverse preparation techniques aiming at improvement of plating efficiency are summarized. Moreover, in view of the outstanding performance, such as corrosion resistance, abrasive resistance and fatigue resistance, this paper critically reviews the behaviors and features of various electroless coatings under different conditions.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

MLEP-Fail calibration for 1/8 inch thick cast plate of 17-4 steel.

Energy Technology Data Exchange (ETDEWEB)

Corona, Edmundo [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2018-02-01

The purpose of the work presented in this memo was to calibrate the Sierra material model Multilinear Elastic-Plastic Hardening Model with Failure (MLEP-Fail) for 1/8 inch thick cast plate of 17-4 steel. The calibration approach is essentially the same as that recently used in a previous memo using data from smooth and notched tensile specimens. The notched specimens were manufactured with three notch radii R = 1=8, 1/32 and 1/64 inches. The dimensions of the smooth and notched specimens are given in the prints in Appendix A. Two cast plates, Plate 3 and Plate 4, with nominally identical properties were considered.

Repairing method for shroud in reactor pressure vessel

International Nuclear Information System (INIS)

Watanabe, Yusuke.

1996-01-01

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

The bipolar plate of AISI 1045 steel with chromized coatings prepared by low-temperature pack cementation for proton exchange membrane fuel cell

Science.gov (United States)

Bai, Ching-Yuan; Wen, Tse-Min; Hou, Kung-Hsu; Ger, Ming-Der

The low-temperature pack chromization, a reforming pack cementation process, is employed to modify AISI 1045 steel for the application of bipolar plates in PEMFC. The process is conducted to yield a coating, containing major Cr-carbides and minor Cr-nitrides, on the substrate in view of enhancing the steel's corrosion resistance and lowering interfacial contact resistance between the bipolar plate and gas diffusion layer. Electrical discharge machining and rolling approach are used as the pretreatment to produce an activated surface on the steel before pack chromization process to reduce operating temperatures and increase deposition rates. The rolled-chromized steel shows the lowest corrosion current density, 3 × 10 -8 A cm -2, and the smallest interfacial contact resistance, 5.9 mΩ cm 2, at 140 N cm -2 among all tested steels. This study clearly states the performance of 1045 carbon steel modified by activated and low-temperature pack chromization processes, which possess the potential to be bipolar plates in the application of PEMFC.

Characteristics of martensite as a function of the M{sub s} temperature in low-carbon armour steel plates

Energy Technology Data Exchange (ETDEWEB)

Maweja, Kasonde, E-mail: mawejak@yahoo.fr [Council for Scientific and Industrial Research, CSIR, Materials Science and Manufacturing, PO Box 395, Pretoria 0001 (South Africa); Department of Materials Science and Metallurgical Engineering, University of Pretoria, Pretoria 0002 (South Africa); Stumpf, Waldo [Department of Materials Science and Metallurgical Engineering, University of Pretoria, Pretoria 0002 (South Africa); Berg, Nic van der [Department of Physics, University of Pretoria, Pretoria 0002 (South Africa)

2009-08-30

The microstructure, morphology, crystal structure and surface relief of martensite in a number of experimental armour steel plates with different M{sub s} temperatures were analysed. Atomic force microscopy, thin foil transmission electron microscopy and scanning electron microscopy allowed the identification of three groups of low-carbon martensitic armour steels. The investigation showed that the size of individual martensite products (plates or packets, laths or blocks) increases as the M{sub s} temperature increases. Comparison of ballistic performances suggests that the morphology (plate or lath) and size of the individual martensite products dictate the effective 'grain size' in resisting fracture or perforation due to ballistic impact.

Microstructure Evolution During Stainless Steel-Copper Vacuum Brazing with a Ag/Cu/Pd Filler Alloy: Effect of Nickel Plating

Science.gov (United States)

Choudhary, R. K.; Laik, A.; Mishra, P.

2017-03-01

Vacuum brazing of stainless steel and copper plates was done using a silver-based filler alloy. In one set of experiments, around 30-µm-thick nickel coatings were electrochemically applied on stainless steel plates before carrying out the brazing runs and its effect in making changes in the braze-zone microstructure was studied. For brazing temperature of 830 °C, scanning electron microscopy examination of the braze-zone revealed that relatively sound joints were obtained when brazing was done with nickel-coated stainless steel than with uncoated one. However, when brazing of nickel-coated stainless steel and copper plates was done at 860 °C, a wide crack appeared in the braze-zone adjacent to copper side. Energy-dispersive x-ray analysis and electron microprobe analysis confirmed that at higher temperature, the diffusion of Cu atoms from copper plate towards the braze-zone was faster than that of Ni atoms from nickel coating. Helium leak rate of the order 10-11 Pa m3/s was obtained for the crack-free joint, whereas this value was higher than 10-4 Pa m3/s for the joint having crack. The shear strength of the joint was found to decrease considerably due to the presence of crack.

Effects of irradiation on the fracture properties of stainless steel weld overlay cladding

International Nuclear Information System (INIS)

Haggag, F.M.; Corwin, W.R.; Nanstad, R.K.

1989-01-01

Stainless steel weld overlay cladding was fabricated using the submerged arc, single-wire, oscillating-electrode, and the three-wire, series-arc methods. Three layers of cladding were applied to a pressure vessel plate to provide adequate thickness for fabrication of test specimens, and irradiations were conducted at temperatures and to fluences relevant to power reactor operation. For the first single-wire method, the first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. The three-wire method used various combinations of types 308, 309, and 304 stainless steel weld wires, and produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. 14 refs., 15 figs., 4 tabs

Effects of implant material and plate design on tendon function and morphology.

Science.gov (United States)

Cohen, Mark S; Turner, Thomas M; Urban, Robert M

2006-04-01

Titanium implants are an alternative to stainless steel implants for internal fixation after fracture. The advantages of titanium include decreased implant stiffness, increased bio-compatibility, and diminished stress shielding. However, titanium has been implicated in tendon irritation and adhesions when used in the hand and wrist. We evaluated the relationship between extensor tendon morphology and dorsal plating of the distal radius in a canine model using distal radius pi plates made of stainless steel, titanium, and titanium alloy with a modified ramped edge design. We found marked histologic changes in the tendons and surrounding soft tissues including tendon deformation and degeneration (fibrillation, cartilage metaplasia, hypocellularity and hyalinization of blood vessels), peritendonous adhesions and neovascularity in the parenchyma. Only a minimal inflammatory cell infiltrate was identified and was limited to the tenosynovium and/or paratenon. No differences were identified between titanium and stainless steel implants and those with a ramped design. Although all animals lost wrist motion with time, no differences were observed between groups. Our results suggest that pi plate placement on the dorsal surface of the distal radius may lead to extensor tendon irritation and dysfunction. There is no evidence to suggest that this is specifically related to titanium or plate edge design.

Experimental study and calculation of boiling heat transfer on steel plates during runout table operation

International Nuclear Information System (INIS)

Liu, Z.D.; Fraser, D.; Samarasekera, I.V.

2002-01-01

Within a hot strip steel mill, red hot steel is hot rolled into a long continuous slab that is led onto what is called the runout table. Temperatures of the steel at the beginning of this table are around 900 o C. Above and below the runout table are banks of water jets, sprays or water curtains that rapidly cool the steel slab. The heat transfer process itself may be considered one of the most complicated in the industrial world. The cooling process that occurs on the runout table is crucial and governs the final mechanical properties and flatness of a steel strip. However, very limited data of industrial conditions has been available and that which is available is poorly understood. To study heat transfer during runout table cooling, an industrial scale pilot runout table facility was constructed at the University of British Columbia (UBC). This paper describes the experimental details, data acquisition and data handling techniques for steel plates during water jet impingement cooling by one circular water jet from industrial headers. The effect of cooling water temperature and initial steel plate temperature as well as varying water jet diameters on heat transfer was systematically investigated. A two-dimensional finite element scheme based inverse heat conduction model was developed to calculate surface heat transfer coefficients along the impinging surface. Heat flux curves at the stagnation area were obtained for selected tests. A quantitative relationship between adjustable processing parameters and heat transfer coefficients along the impinging surface during runout table operation is discussed. The results of the study were used to upgrade an extensive process model developed at UBC. The model ties in the cooling rate and hence two dimensional temperature gradients to the resulting microstructure and final mechanical properties of the steel. This process model is widely used by major steel industries in Canada and the United States. (author)

Prediction of deformations of steel plate by artificial neural network in forming process with induction heating

International Nuclear Information System (INIS)

Nguyen, Truong Thinh; Yang, Young Soo; Bae, Kang Yul; Choi, Sung Nam

2009-01-01

To control a heat source easily in the forming process of steel plate with heating, the electro-magnetic induction process has been used as a substitute of the flame heating process. However, only few studies have analyzed the deformation of a workpiece in the induction heating process by using a mathematical model. This is mainly due to the difficulty of modeling the heat flux from the inductor traveling on the conductive plate during the induction process. In this study, the heat flux distribution over a steel plate during the induction process is first analyzed by a numerical method with the assumption that the process is in a quasi-stationary state around the inductor and also that the heat flux itself greatly depends on the temperature of the workpiece. With the heat flux, heat flow and thermo-mechanical analyses on the plate to obtain deformations during the heating process are then performed with a commercial FEM program for 34 combinations of heating parameters. An artificial neural network is proposed to build a simplified relationship between deformations and heating parameters that can be easily utilized to predict deformations of steel plate with a wide range of heating parameters in the heating process. After its architecture is optimized, the artificial neural network is trained with the deformations obtained from the FEM analyses as outputs and the related heating parameters as inputs. The predicted outputs from the neural network are compared with those of the experiments and the numerical results. They are in good agreement

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

International Nuclear Information System (INIS)

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

1976-12-01

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

Thermo-mechanical treatment of the Cr-Mo constructional steel plates with Nb, Ti and B additions

International Nuclear Information System (INIS)

Adamczyk, J.; Opiela, M.

2002-01-01

Results of investigations of the influence of parameters of thermomechanical treatment, carried out by rolling with controlled recrystallization, on the microstructure and mechanical properties of Cr-Mo constructional steel with Nb, Ti and B microadditions, destined for the manufacturing of weldable heavy plates, are presented. These plates show a yield point of over 960 MPa after heat treatment. Two variants of thermomechanical treatment were worked out, based on the obtained results of investigations, when rolling a plate 40 mm thick in several passes to a plate 15 mm thick in a temperature range from 1100 to 900 o C. It was found that the lack of complete recrystallization of the austenite in the first rolling variant, leads to localization of plastic deformation in form of shear bands. There exists a segregation of MC-type carbides and alloying elements in these bands, causing a distinctive reduction of the crack resistance of the steel, as also a disadvantageous anisotropy of plastic properties of plate after tempering. For plates rolled under the same conditions, using a retention shield, a nearly three times higher impact energy in - 40 o C was obtained, as also only a slight anisotropy of plastic properties, saving the required mechanical properties. (author)

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.)

Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

International Nuclear Information System (INIS)

Nanstad, R.K.

1980-01-01

The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior is characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens

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

International Nuclear Information System (INIS)

Ebine, N.; Suzuki, M.

2001-01-01

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

Laser cutting of steel plates up to 100 mm in thickness with a 6-kW fiber laser for application to dismantling of nuclear facilities

Science.gov (United States)

Shin, Jae Sung; Oh, Seong Yong; Park, Hyunmin; Chung, Chin-Man; Seon, Sangwoo; Kim, Taek-Soo; Lee, Lim; Lee, Jonghwan

2018-01-01

A cutting study with a high-power ytterbium-doped fiber laser was conducted for the dismantling of nuclear facilities. Stainless steel and carbon steel plates of various thicknesses were cut at a laser power of 6-kW. Despite the use of a low output of 6-kW, the cutting was successful for both stainless steel and carbon steel plates of up to 100 mm in thickness. In addition, the maximum cutting speeds against the thicknesses were obtained to evaluate the cutting performance. As representative results, the maximum cutting speeds for a 60-mm thickness were 72 mm/min for the stainless steel plates and 35 mm/min for the carbon steel plates, and those for a 100-mm thickness were 7 mm/min for stainless steel and 5 mm/min for carbon steel plates. These results show an efficient cutting capability of about 16.7 mm by kW, whereas other groups have shown cutting capabilities of ∼10 mm by kW. Moreover, the maximum cutting speeds were faster for the same thicknesses than those from other groups. In addition, the kerf widths of 60-mm and 100-mm thick steels were also obtained as another important parameter determining the amount of secondary waste. The front kerf widths were ∼1.0 mm and the rear kerf widths were larger than the front kerf widths but as small as a few millimeters.

Characterization of morphology and kinetics of bainite transformation in a low alloy steel

International Nuclear Information System (INIS)

Gupta, C.; Dey, G.K.; Srivastav, D.; Chakravarthy, J.K.; Banerjee, S.

2005-01-01

Bainite transformation is ubiquitous in steels for pressure vessel applications in thermal and nuclear power plants. In this class of steels bainite is the dominant phase found in the microstructure, after industrial thermo-mechanical processing and heat treatment of pressure vessel component. The study of bainite transformation has been carried out using both isothermal and continuous cooling conditions. Previous studies have reported significant differences in the morphology and the type of bainite formed under these two conditions. Continuous cooling has been shown to result in a wider variety of bainite transformation products as compared with isothermal treatments. This has important implications for the technological properties of power plant components such as strength, toughness and hardenability. In the present study the cooling transformation characteristics of a new CrMo pressure vessel steel has been examined using dilatometry supplemented with TEM examination. The dilatometric data were analyzed to determine the activation energy and Avrami exponents. It was found that bainite with different morphologies formed over the cooling rates employed and were kinetically distinct. The dilatometric study along with TEM studies has shown that non-isothermal decomposition of austenite in this steel results in a complex microstructure containing an array of bainite morphologies. The bainitic ferrite plates are seen to be associated with various inter- and intra- plate constituents as the cooling rate changes. Despite this the transformation remains essentially bainitic over the range of cooling rates studied. Three different cooling rate regimes with distinctly different calculated Avrami exponents have been observed. (author)

Tailoring diffraction technique Rietveld method on residual stress measurements of cold-can oiled 304 stainless steel plates

International Nuclear Information System (INIS)

Parikin; Killen, P.; Anis, M.

2003-01-01

Tailoring of diffraction technique-Rietveld method on residual stress measurements of cold-canailed stainless steel 304 plates assuming the material is isotopic, the residual stress measurements using X-ray powder diffraction is just performed for a plane lying in a large angle. For anisotropic materials, the real measurements will not be represented by the methods. By Utilizing of all diffraction peaks in the observation region, tailoring diffraction technique-Rietveld analysis is able to cover the limitations. The residual stress measurement using X-ray powder diffraction tailored by Rietveld method, in a series of cold-canailed stainless steel 304 plates deforming; 0, 34, 84, 152, 158, 175, and 196 % reduction in thickness, have been reported. The diffraction data were analyzed by using Rietveld structure refinement method. Also, for all cold-canailed stainless steel 304 plates cuplikans, the diffraction peaks are broader than the uncanailed one, indicating that the strains in these cuplikans are inhomogeneous. From an analysis of the refined peak shape parameters, the average root-mean square strain, which describes the distribution of the inhomogeneous strain field, was calculated. Finally, the average residual stresses in cold-canailed stainless steel 304 plates were shown to be a combination effect of hydrostatic stresses of martensite particles and austenite matrix. The average residual stresses were evaluated from the experimentally determined average lattice strains in each phase. It was found the tensile residual stress in a cuplikan was maximum, reaching 442 MPa, for a cuplikan reducing 34% in thickness and minimum for a 196% cuplikan

Digital image rectification tool for metrification of gusset plate connections in steel truss bridges.

Science.gov (United States)

2009-03-01

A method was developed to obtain dimensional data from photographs for analyzing steel truss gusset plate : connections. The method relies on a software application to correct photographic distortion and to scale the : photographs for analysis. The a...

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

International Nuclear Information System (INIS)

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

1991-01-01

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

The bipolar plate of AISI 1045 steel with chromized coatings prepared by low-temperature pack cementation for proton exchange membrane fuel cell

Energy Technology Data Exchange (ETDEWEB)

Bai, Ching-Yuan; Ger, Ming-Der [Department of Chemistry and Materials Science and Engineering, Chung Cheng Institute of Technology, National Defense University, Tau-Yuan 335 (China); Wen, Tse-Min [School of Defense Science, Chung Cheng Institute of Technology, National Defense University, Tao-Yuan 335 (China); Hou, Kung-Hsu [Department of Power Vehicles and System Engineering, Chung Cheng Institute of Technology, National Defense University, Tao-Yuan 335 (China)

2010-02-01

The low-temperature pack chromization, a reforming pack cementation process, is employed to modify AISI 1045 steel for the application of bipolar plates in PEMFC. The process is conducted to yield a coating, containing major Cr-carbides and minor Cr-nitrides, on the substrate in view of enhancing the steel's corrosion resistance and lowering interfacial contact resistance between the bipolar plate and gas diffusion layer. Electrical discharge machining and rolling approach are used as the pretreatment to produce an activated surface on the steel before pack chromization process to reduce operating temperatures and increase deposition rates. The rolled-chromized steel shows the lowest corrosion current density, 3 x 10{sup -8} A cm{sup -2}, and the smallest interfacial contact resistance, 5.9 m{omega} cm{sup 2}, at 140 N cm{sup -2} among all tested steels. This study clearly states the performance of 1045 carbon steel modified by activated and low-temperature pack chromization processes, which possess the potential to be bipolar plates in the application of PEMFC. (author)

Learning from EDF investigations on SG divider plates and vessel head nozzles. Evidence of prior deformation effect on stress corrosion cracking

International Nuclear Information System (INIS)

Deforge, D.; Duisabeau, L.; Miloudi, S.; Thebault, Y.; Couvant, T.; Vaillant, F.; Lemaire, E.

2011-01-01

Nickel Based alloys Stress Corrosion Cracking (SCC) has been a major concern for all the Nuclear Power Plants (NPP) utilities since the beginning of the seventies. At EDF, the nineties were marked by the occurrence of cracks on vessel head nozzles. These cracks were responsible for a leak at Bugey 3 vessel head, which was the precursor leading to the replacement of all vessel heads. From 2002, new cases of Stress Corrosion Cracking were reported on Steam Generator (SG) Divider Plates (SGDP) welded junctions. These cracks are periodically inspected inservice and reparations could be performed in case of a significant evolution of the phenomenon even if the safety issue is less relevant than for the vessel head nozzles. Both issues have led to an important non-destructive testing (NDT) program and to destructive investigations campaigns. NDT were performed on an exhaustive basis for all vessel head nozzles and for all the divider plates of 900 MWe plants. Destructive investigations were performed on more than 30 vessel head nozzles and on 6 divider plates. The last investigations were performed on samples from two decommissioned Steam Generators of Chinon B1 which present SCC cracks. In this paper, the main conclusions driven from the analysis of both NDT and destructive investigation results are reported and a comparison of the behaviours of divider plates and vessel head nozzles is given. Results give evidence that prior plastic deformation of the components before operation is fundamental for the further environmental behaviour of the material. Analysis of field experience based on parameters characteristics of prior deformation and parameters characteristics of material microstructure can be used to account for the components which are the most sensitive to SCC cracking. Some perspectives on SCC predictive models are also presented. (authors)

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

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

International Nuclear Information System (INIS)

Shewmon, P.G.

1980-12-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

The relation between the amount of dissolved water and metals dissolved from stainless steel or aluminum plate in safflower oil

International Nuclear Information System (INIS)

Takasago, Masahisa; Takaoka, Kyo

1986-01-01

The amount of water dissolved in safflower oil at the frying temperature (180 deg C) was 518 ∼ 1012 ppm, allowing water to drop continuously (0.035 g/2 min) into the oil for 1 ∼ 3 h. When the oil was heated with metal plates under the same conditions, the amount of dissolved water in the oil increased more than in the absence of the metal plates. In case of stainless steel, the amount was 1.26 to 1.33 times, and with aluminum plates, 1.06 to 1.13 times the amount without plates. When these metal plates were heated with the oil under the above conditions, the water dissolved the metal of the plates into the oil. In case of stainless steel, iron dissolved from 0.17 to 0.77 ppm, nickel, 0.04 ppm and chromium, from 0.02 to 0.03 ppm. Similarly, the amount of aluminum dissolved from the aluminum plate was from 0.10 to 0.45 ppm. (author)

Relation between the amount of dissolved water and metals dissolved from stainless steel or aluminum plate in safflower oil

Energy Technology Data Exchange (ETDEWEB)

Takasago, Masahisa; Takaoka, Kyo

1986-12-01

The amount of water dissolved in safflower oil at the frying temperature (180 deg C) was 518 -- 1012 ppM, allowing water to drop continuously (0.035 g/2 min) into the oil for 1 -- 3 h. When the oil was heated with metal plates under the same conditions, the amount of dissolved water in the oil increased more than in the absence of the metal plates. In case of stainless steel, the amount was 1.26 to 1.33 times, and with aluminum plates, 1.06 to 1.13 times the amount without plates. When these metal plates were heated with the oil under the above conditions, the water dissolved the metal of the plates into the oil. In case of stainless steel, iron dissolved from 0.17 to 0.77 ppM, nickel, 0.04 ppM and chromium, from 0.02 to 0.03 ppM. Similarly, the amount of aluminum dissolved from the aluminum plate was from 0.10 to 0.45 ppM.

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

International Nuclear Information System (INIS)

Rao, S.

1980-09-01

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

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

International Nuclear Information System (INIS)

Holt, J.; Goddard, D.J.

1980-01-01

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

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

International Nuclear Information System (INIS)

Rauscher, F.

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Rauscher, F

2003-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1967-09-15

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

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

International Nuclear Information System (INIS)

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

1967-09-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

An introduction to the PISC II project - programme for the inspection of steel components

International Nuclear Information System (INIS)

Nichols, R.; McDonald, N.R.

1987-01-01

The paper describes the work of the Plate Inspection Steering Committee (PISC) on the non-destructive examination of reactor pressure vessel steels. A description is given of the PISC I exercise on flaw measurements in test plates, including the PISC procedure and the alternative procedures in the PISC I exercise. The motivation for a PISC II programme is described, together with the objectives and terms of reference of PISC II. (U.K.)

Ductile fracture toughness of modified A 302 grade B plate materials. Volume 2

International Nuclear Information System (INIS)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-02-01

The objective of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A 302 grade B plate materials typical of those used in fabricating reactor pressure vessels. A previous experimental study at Materials Engineering Associates (MEA) on one particular heat of A 302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in numerous tests made on the more recent production materials of A 533 grade B and A 508 class 2 pressure vessel steels. It was unknown if the departure from norm for the MEA material was a generic characteristic for all heats of A 302 grade B steels or just unique to that one particular plate. Seven heats of modified A 302 grade B steel and one heat of vintage A 533 grade B steel were provided to this project by the General Electric Company of San Jose, California. All plates were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550 degrees F (288 degrees C). Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, IT, 2T, and 4T). None of the seven heats of modified A 302 grade showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550 degrees F (82 to 288 degrees C) produced the usual loss in J-R curve fracture toughness. Generic J-R curves and mathematical curve fits to the same were generated to represent each heat of material. This volume is a compilation of all data developed

Aluminum and Other Coatings for the Passivation of Tritium Storage Vessels

Energy Technology Data Exchange (ETDEWEB)

Spencer, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Korinko, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-11-16

Using a highly sensitive residual gas analyzer, the off-gassing of hydrogen, water, and hydrocarbons from surface-treated storage vessels containing deuterium was measured. The experimental storage vessels were compared to a low-off-gassing, electro-polished 304L canister. Alternative vessels were made out of aluminum, or were coatings on 304L steel. Coatings included powder pack aluminide, electro-plated aluminum, powder pack chromide, dense electro-plated chromium, copper plated, and copper plated with 25 and 50 percent nano-diamond. Vessels were loaded with low pressure deuterium to observe exchange with protium or hydrogen as observed with formation of HD and HDO. Off gas of D₂O or possible CD₄ was observed at mass 20. The main off-gas in all of the studies was H₂. The studies indicated that coatings required significant post-coating treatment to reduce off-gas and enhance the permeation barrier from gases likely added during the coating process. Dense packed aluminum coatings needed heating to drive off water. Electro-plated aluminum, chromium and copper coatings appeared to trap hydrogen from the plating process. Nano-diamond appeared to enhance the exchange rate with hydrogen off gas, and its coating process trapped significant amounts of hydrogen. Aluminum caused more protium exchange than chromium-treated surfaces. Aluminum coatings released more water, but pure aluminum vessels released small amounts of hydrogen, little water, and generally performed well. Chromium coating had residual hydrogen that was difficult to totally outgas but otherwise gave low residuals for water and hydrocarbons. Our studies indicated that simple coating of as received 304L metal will not adequately block hydrogen. The base vessel needs to be carefully out-gassed before applying a coating, and the coating process will likely add additional hydrogen that must be removed. Initial simple bake-out and leak checks up to 350° C for a few hours was

Improved performance of brazed plate heat exchangers made of stainless steel type EN 1.4401 (UNS S31600) when using a iron-based braze filler

Energy Technology Data Exchange (ETDEWEB)

Sjoedin, P. [Alfa Laval Materials, Lund (Sweden)

2004-07-01

The mechanical properties of brazed plate heat exchangers, made of stainless steel plates type EN 1.4401, brazed with a new iron-based braze filler ''AlfaNova'', have been evaluated. The results were compared with heat exchangers brazed with a copper (pure copper) and a nickel-based (MBF 51) braze filler. Their resistance against pressure- and temperature fatigue, which are important for the lifetime of a heat exchanger, and the burst pressure, which is important for pressure vessel approvals, were tested and evaluated. It was found that the pressure fatigue resistance was extraordinary good for the heat exchangers brazed the iron-based filler and its temperature fatigue resistance was better than those brazed with nickel-based braze filler and slightly lower than those brazed with copper. The highest burst pressures were achieved for the copper brazed units followed by the iron-brazed units and rearmost the nickel-brazed units. (orig.)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Ag-polytetrafluoroethylene composite coating on stainless steel as bipolar plate of proton exchange membrane fuel cell

Energy Technology Data Exchange (ETDEWEB)

Fu, Yu. [Laboratory of Fuel Cells, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, Zhongshan Road, Dalian 116023 (China); Graduate University of Chinese Academy of Sciences, Beijing 100049 (China); Hou, Ming; Shao, Zhigang; Yi, Baolian [Laboratory of Fuel Cells, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, Zhongshan Road, Dalian 116023 (China); Xu, Hongfeng; Hou, Zhongjun; Ming, Pingwen [Sunrise Power Co., Ltd., Dalian 116025 (China)

2008-08-01

Forming a coating on metals by surface treatment is a good way to get high performance bipolar plate of proton exchange membrane fuel cell (PEMFC). In our research, Ag-polytetrafluoroethylene (PTFE) composite film was electrodeposited with silver-gilt solution of nicotinic acid by a bi-pulse electroplating power supply on 316 L stainless steel bipolar plate of PEMFC. Surface topography, contact angle, interfacial conductivity and corrosion resistance of the bipolar plate samples were investigated. Results showed that the defects on the Ag-PTFE composite coating are greatly reduced compared with those on the pure Ag coating fabricated under the same condition; and the contact angle of the Ag-PTFE composite coating with water is 114 , which is much bigger than that of the pure Ag coating (73 ). In addition, the interfacial contact resistance of the composite coating stays as low as the pure Ag coating; and the bipolar plate sample with composite coating shows a close corrosion resistance to the pure Ag coating sample in potentiodynamic and potentiostatic tests. Coated 316 L stainless steel plate with Ag-PTFE composite coating exhibits well hydrophobic characteristic, less defects, high interfacial conductivity and good corrosion resistance, which shows a great potential of the application in PEMFC. (author)

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

International Nuclear Information System (INIS)

Van Der Sluys, W.A.

1996-10-01

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

Guide to the periodic inspection of nuclear reactor steel pressure vessels

International Nuclear Information System (INIS)

1969-01-01

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

Processing vessel for high level radioactive wastes

International Nuclear Information System (INIS)

Maekawa, Hiromichi

1998-01-01

Upon transferring an overpack having canisters containing high level radioactive wastes sealed therein and burying it into an underground processing hole, an outer shell vessel comprising a steel plate to be fit and contained in the processing hole is formed. A bury-back layer made of dug earth and sand which had been discharged upon forming the processing hole is formed on the inner circumferential wall of the outer shell vessel. A buffer layer having a predetermined thickness is formed on the inner side of the bury-back layer, and the overpack is contained in the hollow portion surrounded by the layer. The opened upper portion of the hollow portion is covered with the buffer layer and the bury-back layer. Since the processing vessel having a shielding performance previously formed on the ground, the state of packing can be observed. In addition, since an operator can directly operates upon transportation and burying of the high level radioactive wastes, remote control is no more necessary. (T.M.)

Study on in-vessel ISI for JOYO. Ultrasound propagation characteristic in the core support plate

International Nuclear Information System (INIS)

Ariyoshi, Masahiko; Ara, Kuniaki; Hirabayashi, Masaru

2005-03-01

The report describes the feasibility study on the in-vessel inspection technique to be applied for the experimental fast reactor JOYO. The object of this examination is to confirm the integration of reactor structure under sodium environment by an immediate means. The core support plate which is an important structure supports the weight of the core assembly is selected to an object of the inspection. In the examination until last year, the core support plate inspection equipment concept which combined ultrasound sensor with manipulator was constructed. In this concept, the ultrasound sensor is accessed to a low-pressure plenum sidewall and integrity of the core support plate weld is inspected. In this study, the ultrasound propagation behavior was examined to confirm the range where the core support plate by this concept was able to be inspected. The outline result is shown follows. (1) Only the transverse wave can be generated in the structure material by reflecting the incidence longitudinal wave from the sensor in the wedge. The use of this transverse wave is effective in the core support plate inspection. (2) Because the attenuation of the ultrasound wave depends on the distance, the sensor is made to approach from the fuel rack in the reactor vessel about two places in the upper part of the core support plate weld far from low-pressure plenum. (3) It is necessary to evaluate the permeability of the ultrasound wave by the mock-up examination in consideration of a peculiar attenuation of the structural material, the reflectivity from defect, etc. (4) In the core support plate inspection of phenix reactor, a weld about 4m away from the sensor position is inspected by using the Lamb wave. In this inspection, because it was generated to echo according to the geometrical shape of the structure material, the evaluation method by the analysis to identify the echo from the defect was constructed, and it was verified by the mock-up examination. It is preferable that

Evaluation of AISI 316L stainless steel welded plates in heavy petroleum environment

International Nuclear Information System (INIS)

Carvalho Silva, Cleiton; Pereira Farias, Jesualdo; Batista de Sant'Ana, Hosiberto

2009-01-01

This work presents the study done on the effect of welding heating cycle on AISI 316L austenitic stainless steel corrosion resistance in a medium containing Brazilian heavy petroleum. AISI 316L stainless steel plates were welded using three levels of welding heat input. Thermal treatments were carried out at two levels of temperatures (200 and 300 deg. C). The period of treatment in all the trials was 30 h. Scanning electronic microscopy (SEM) and analysis of X-rays dispersive energy (EDX) were used to characterize the samples. Weight loss was evaluated to determine the corrosion rate. The results show that welding heating cycle is sufficient to cause susceptibility to corrosion caused by heavy petroleum to the heat affected zone (HAZ) of the AISI 316L austenitic stainless steel

Difference in metallic wear distribution released from commercially pure titanium compared with stainless steel plates.

Science.gov (United States)

Krischak, G D; Gebhard, F; Mohr, W; Krivan, V; Ignatius, A; Beck, A; Wachter, N J; Reuter, P; Arand, M; Kinzl, L; Claes, L E

2004-03-01

Stainless steel and commercially pure titanium are widely used materials in orthopedic implants. However, it is still being controversially discussed whether there are significant differences in tissue reaction and metallic release, which should result in a recommendation for preferred use in clinical practice. A comparative study was performed using 14 stainless steel and 8 commercially pure titanium plates retrieved after a 12-month implantation period. To avoid contamination of the tissue with the elements under investigation, surgical instruments made of zirconium dioxide were used. The tissue samples were analyzed histologically and by inductively coupled plasma atomic emission spectrometry (ICP-AES) for accumulation of the metals Fe, Cr, Mo, Ni, and Ti in the local tissues. Implant corrosion was determined by the use of scanning electron microscopy (SEM). With grades 2 or higher in 9 implants, steel plates revealed a higher extent of corrosion in the SEM compared with titanium, where only one implant showed corrosion grade 2. Metal uptake of all measured ions (Fe, Cr, Mo, Ni) was significantly increased after stainless steel implantation, whereas titanium revealed only high concentrations for Ti. For the two implant materials, a different distribution of the accumulated metals was found by histological examination. Whereas specimens after steel implantation revealed a diffuse siderosis of connective tissue cells, those after titanium exhibited occasionally a focal siderosis due to implantation-associated bleeding. Neither titanium- nor stainless steel-loaded tissues revealed any signs of foreign-body reaction. We conclude from the increased release of toxic, allergic, and potentially carcinogenic ions adjacent to stainless steel that commercially pure Ti should be treated as the preferred material for osteosyntheses if a removal of the implant is not intended. However, neither material provoked a foreign-body reaction in the local tissues, thus cpTi cannot be

76 FR 25666 - Stainless Steel Plate in Coils from Belgium: Final Results of Full Sunset Review and Revocation...

Science.gov (United States)

2011-05-05

... Corporation and the United Steel, Paper and Forestry, Rubber, Manufacturing, Energy, Allied Industrial and... included in the scope of the AD orders on SSPC from Belgium, Italy, South Africa, the Republic of Korea... Certain Stainless Steel Plate in Coils From Belgium, Italy, South Korea, South Africa, and Taiwan, and the...

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)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

2005-06-01

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

76 FR 50495 - Stainless Steel Plate From Belgium, Italy, Korea, South Africa, and Taiwan

Science.gov (United States)

2011-08-15

... Review] Stainless Steel Plate From Belgium, Italy, Korea, South Africa, and Taiwan Determinations On the... Africa, and Taiwan would be likely to lead to continuation or recurrence of material injury to an industry in the United States within a reasonably foreseeable time.\\2\\ The Commission further determines...

A preliminary study on the local impact behavior of Steel-plate Concrete walls

International Nuclear Information System (INIS)

Kim, Kap-sun; Moon, Il-hwan; Choi, Hyung-jin; Nam, Deok-woo

2017-01-01

International regulations for nuclear power plants strictly prescribe the design requirements for local impact loads, such as aircraft engine impact, and internal and external missile impact. However, the local impact characteristics of Steel-plate Concrete (SC) walls are not easy to evaluate precisely because the dynamic impact behavior of SC walls which include external steel plate, internal concrete, tie-bars, and studs, is so complex. In this study, dynamic impact characteristics of SC walls subjected to local missile impact load are investigated via actual high-speed impact test and numerical simulation. Three velocity checkout tests and four SC wall tests were performed at the Energetic Materials Research and Testing Center (EMRTC) site in the USA. Initial and residual velocity of the missile, strain and acceleration of the back plate, local failure mode (penetration, bulging, splitting and perforation) and deformation size, etc. were measured to study the local behavior of the specimen using high speed cameras and various other instrumentation devices. In addition, a more advanced and applicable numerical simulation method using the finite element (FE) method is proposed and verified by the experimental results. Finally, the experimental results are compared with the local failure evaluation formula for SC walls recently proposed, and future research directions for the development of a refined design method for SC walls are reviewed.

Radioactive material-containing vessel and method of manufacturing the same

International Nuclear Information System (INIS)

Kanazawa, Hiroshi; Wada, Katsuyoshi; Ota, Shigeo; Nishioka, Eiji; Okuno, Michinori.

1995-01-01

In a vessel for containing radioactive materials having an outer wall with a structure of interposing a lead layer, as a shielding material between inner and outer cylinders made of steel plates, the inner cylinder and the lead layer are in close contact by way of a thin layer of a lead/tin type soldering material and to such an extent that the boundary layer is not detected by supersonic inspection. In addition, flux is coated to the steel plate, which forms the inner cylinder, on the surface being in contact with the lead layer, then a thin layer of the soldering material such as lead or tin is formed, to cast the lead between the inner and the outer cylinders. Then, since the inner cylinder and the lead layer are thermally joined tightly, heat generated at the inside can effectively be released to the outside, so that it is effective as a high-performance cask for transporting a large amount of radioactive materials such as spent nuclear fuels having high temperature afterheat. In addition, a containing vessel with good contact between the inner cylinder and the lead can be manufactured at a low cost only applying a simple primer treatment on the surface of the inner cylinder in addition to an existent lead casting method. (N.H.)

Treatment of Industrial Liquid Waste of Steel Plating by Coagulation-Flocculation Using Sodium Biphosphate

International Nuclear Information System (INIS)

Subiarto; Herlan Martono

2007-01-01

Research about treatment of industrial liquid waste of steel plating by coagulation-flocculation using sodium biphosphate have been conducted. The purpose of the treatment was the content reduction of Cr, Ni, and Cu in the liquid waste, so that produced effluent with Cr, Ni, and Cu content until they laid under mutual standard. The variables studied in this process were the solution pH, the coagulant/waste volume comparison, the speed of the fast stirring, and the time of the fast stirring. Optimum separation efficiency on coagulation-flocculation process of liquid waste of steel plating using sodium biphosphate at the condition of solution ph 9, coagulant/waste volume comparation 1.50, the speed of the fast stirring 400 rpm, and the time of fast stirring is 5 minute. Low stirring was conducted at 60 rpm for 60 minute. The yields of optimum separation efficiency in this condition were 99.48 % for Cr, 99.51 % for Ni, and 99.03 % for Cu. (author)

Hydrogen permeation inhibition by zinc-nickel alloy plating on steel XC68

International Nuclear Information System (INIS)

El Hajjami, A.; Gigandet, M.P.; De Petris-Wery, M.; Catonne, J.C.; Duprat, J.J.; Thiery, L.; Raulin, F.; Starck, B.; Remy, P.

2008-01-01

The inhibition of hydrogen permeation and barrier effect by zinc-nickel plating was investigated using the Devanathan-Stachurski permeation technique. The hydrogen permeation and hydrogen diffusion for the zinc-nickel (12-15%) plating on steel XC68 is compared with zinc and nickel. Hydrogen permeation and hydrogen diffusion were followed as functions of time at current density applied (cathodic side) and potential permanent (anodic side). The hydrogen permeation inhibition for zinc-nickel is intermediate to that of nickel and zinc. This inhibition was due to nickel-rich layer effects at the Zn-Ni alloy/substrate interface, is shown by GDOES. Zinc-nickel plating inhibited the hydrogen diffusion greater as compared to zinc. This diffusion resistance was due to the barrier effect caused by the nickel which is present at the interface and transformed the hydrogen atomic to Ni 2 H compound, as shown by GIXRD.

Hydrogen permeation inhibition by zinc-nickel alloy plating on steel XC68

Energy Technology Data Exchange (ETDEWEB)

El Hajjami, A. [Institut UTINAM, UMR CNRS 6213, Sonochimie et Reactivite des Surfaces, Universite de Franche-Comte, 16 route de Gray, 25030 Besancon Cedex (France); Coventya S.A.S., 51 rue Pierre, 92588 Clichy Cedex (France); Gigandet, M.P. [Institut UTINAM, UMR CNRS 6213, Sonochimie et Reactivite des Surfaces, Universite de Franche-Comte, 16 route de Gray, 25030 Besancon Cedex (France)], E-mail: marie-pierre.gigandet@univ-fcomte.fr; De Petris-Wery, M. [Institut Universitaire de Technologie d' Orsay, Universite Paris XI, Plateau de Moulon, 91400 Orsay (France); Catonne, J.C. [Professeur Honoraire du Conservatoire national des arts et metiers (CNAM), Paris (France); Duprat, J.J.; Thiery, L.; Raulin, F. [Coventya S.A.S., 51 rue Pierre, 92588 Clichy Cedex (France); Starck, B.; Remy, P. [Lisi Automotive, 28 faubourg de Belfort, BP 19, 90101 Delle Cedex (France)

2008-12-30

The inhibition of hydrogen permeation and barrier effect by zinc-nickel plating was investigated using the Devanathan-Stachurski permeation technique. The hydrogen permeation and hydrogen diffusion for the zinc-nickel (12-15%) plating on steel XC68 is compared with zinc and nickel. Hydrogen permeation and hydrogen diffusion were followed as functions of time at current density applied (cathodic side) and potential permanent (anodic side). The hydrogen permeation inhibition for zinc-nickel is intermediate to that of nickel and zinc. This inhibition was due to nickel-rich layer effects at the Zn-Ni alloy/substrate interface, is shown by GDOES. Zinc-nickel plating inhibited the hydrogen diffusion greater as compared to zinc. This diffusion resistance was due to the barrier effect caused by the nickel which is present at the interface and transformed the hydrogen atomic to Ni{sub 2}H compound, as shown by GIXRD.

Flexural strength using Steel Plate, Carbon Fiber Reinforced Polymer (CFRP) and Glass Fiber Reinforced Polymer (GFRP) on reinforced concrete beam in building technology

Science.gov (United States)

Tarigan, Johannes; Patra, Fadel Muhammad; Sitorus, Torang

2018-03-01

Reinforced concrete structures are very commonly used in buildings because they are cheaper than the steel structures. But in reality, many concrete structures are damaged, so there are several ways to overcome this problem, by providing reinforcement with Fiber Reinforced Polymer (FRP) and reinforcement with steel plates. Each type of reinforcements has its advantages and disadvantages. In this study, researchers discuss the comparison between flexural strength of reinforced concrete beam using steel plates and Fiber Reinforced Polymer (FRP). In this case, the researchers use Carbon Fiber Reinforced Polymer (CFRP) and Glass Fiber Reinforced Polymer (GFRP) as external reinforcements. The dimension of the beams is 15 x 25 cm with the length of 320 cm. Based on the analytical results, the strength of the beam with CFRP is 1.991 times its initial, GFRP is 1.877 times while with the steel plate is 1.646 times. Based on test results, the strength of the beam with CFRP is 1.444 times its initial, GFRP is 1.333 times while the steel plate is 1.167 times. Based on these test results, the authors conclude that beam with CFRP is the best choice for external reinforcement in building technology than the others.

Tribology aspects of a pressure vessel closure subjected to pressure cycling

International Nuclear Information System (INIS)

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

1988-04-01

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

FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

International Nuclear Information System (INIS)

Altstadt, E.; Willschuetz, H.G.; Mueller, G.

2004-01-01

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)

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.

Laser cut hole matrices in novel armour plate steel for appliqué battlefield vehicle protection

OpenAIRE

Thomas, Daniel J.

2016-01-01

During this research, experimental rolled homogeneous armour steel was cast, annealed and laser cut to form an appliqué plate. This Martensitic–Bainitic microstructure steel grade was used to test a novel means of engineering lightweight armour. It was determined that a laser cutting speed of 1200 mm/min produced optimum hole formations with limited distortion. The array of holes acts as a double-edged solution, in that they provide weight saving of 45%, providing a protective advantage and i...

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Revision of the fracture models in steels for nuclear pressure vessels

Energy Technology Data Exchange (ETDEWEB)

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

1981-01-01

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

Corrosion kinetics of 316L stainless steel bipolar plate with chromiumcarbide coating in simulated PEMFC cathodic environment

Directory of Open Access Journals (Sweden)

N.B. Huang

Full Text Available Stainless steel with chromium carbide coating is an ideal candidate for bipolar plates. However, the coating still cannot resist the corrosion of a proton exchange membrane fuel cell (PEMFC environment. In this work, the corrosion kinetics of 316L stainless steel with chromium carbide is investigated in simulated PEMFC cathodic environment by combining electrochemical tests with morphology and microstructure analysis. SEM results reveal that the steel’s surface is completely coated by Cr and chromium carbide but there are pinholes in the coating. After the coated 316L stainless steel is polarized, the diffraction peak of Fe oxide is found. EIS results indicate that the capacitive resistance and the reaction resistance first slowly decrease (2–32 h and then increase. The potentiostatic transient curve declines sharply within 2000 s and then decreases slightly. The pinholes, which exist in the coating, result in pitting corrosion. The corrosion kinetics of the coated 316L stainless steel are modeled and accords the following equation: i0 = 7.6341t−0.5, with the corrosion rate controlled by ion migration in the pinholes. Keywords: PEMFC, Metal bipolar plate, Chromium carbide coating, Corrosion kinetics, Pitting corrosion

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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.

The effect of filler metal thickness on residual stress and creep for stainless-steel plate-fin structure

Energy Technology Data Exchange (ETDEWEB)

Jiang Wenchun [School of Mechanical and Power Engineering, Nanjing University of Technology, Nanjing 210009 (China)], E-mail: jiangwenchun@126.com; Gong Jianming; Chen Hu; Tu, S.T. [School of Mechanical and Power Engineering, Nanjing University of Technology, Nanjing 210009 (China)

2008-08-15

Stainless-steel plate-fin heat exchanger (PFHE) has been used as a high-temperature recuperator in microturbine for its excellent qualities in compact structure, high-temperature and pressure resistance. Plate-fin structure, as the core of PFHE, is fabricated by vacuum brazing. The main component fins and the parting sheets are joined by fusion of a brazing alloy cladded to the surface of parting sheets. Owing to the material mismatching between the filler metal and the base metal, residual stresses can arise and decrease the structure strength greatly. The recuperator serves at high temperature and the creep would happen. The thickness of the filler metal plays an important role in the joint strength. Hence this paper presented a finite element (FE) analysis of the brazed residual stresses and creep for a counterflow stainless-steel plate-fin structure. The effect of the filler metal thickness on residual stress and creep was investigated, which provides a reference for strength design.

Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

Energy Technology Data Exchange (ETDEWEB)

R.M. Dolin

2003-03-03

This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

Supply chain of steel industries for the nuclear power plant construction in Indonesia

International Nuclear Information System (INIS)

Dharu Dewi; Sahala M Lumbanraja

2017-01-01

Nuclear Power Plant (NPP) Construction needs steel materials for the manufacturing of heavy components and civil work construction. National industries is expected to supply steel components especially for non nuclear component needs. Supply chain of steel industries is required to know the potency of steel industries from upstream to downstream industries which can support the NPP construction sustainability. The type of steel needed in the NPP construction consist of structure steel, rebar, steel plate, etc. The aim of the study is to identify supply chain of steel industries from upstream industries to downstream industries so that they can supply steel needs in the NPP construction. The methodology used are literature review and industries survey by purposive sampling test which sent questionnaires and carrying out technical visits to the potential industries to supply steel components for NPP construction. From the analysis of the questionnaires and survey, it has been obtained that the Indonesian steel industries capable of supplying steel for construction materials of non-nuclear parts are PT. Krakatau Steel, PT. Gunung Steel Group (PT Gunung Garuda and PT. Gunung Raja Paksi), PT. Cilegon Fabricators and PT. Ometraco Arya Samanta. While steel materials for primary components with nuclear grade, such as steel materials for reactor vessels and pressure vessels, the Indonesian steel industry has not been able to supply them. Therefore, the Indonesian steel industries must improve its capability, both in raw material processing and fabrication capability in order to meet the requirements of specifications, codes and standards of nuclear grade. (author)

Dutch supplier rewarded for manufacture of the two vacuum vessels for the ATLAS end-cap toroids

CERN Multimedia

Maximilien Brice

2003-01-01

The ATLAS collaboration has presented an award for outstanding supplier performance to Dutch firm Schelde Exotech. Based on a design by Rutherford Appleton Laboratory, UK, Schelde Exotech manufactured under a NIKHEF contract the two 500 m3 large vacuum vessels for the cryostats of the ATLAS end-cap toroids. These 11-metre diameter castellated aluminium vessels with stainless-steel bore tube are essentially made up of 40-mm-thick plates for the shells, 75-mm-thick plates for the endplates, and 150-mm-thick bars for the flanges. Because of transport constraints, the vessels were made in halves, temporarily sealed and vacuum tested at the works, then transported to CERN for final assembly and acceptance tests. Both vessels were vacuum-tight and the meticulous and clean way of working ensured that a high vacuum was obtained within a few days of pumping. The delivery to CERN was completed in July 2002. Representatives of Schelde Exotech are seen here receiving their award in the ATLAS assembly hall. In the backgro...

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

The effect of electrode vertex angle on automatic tungsten-inert-gas welds for stainless steel 304L plates

International Nuclear Information System (INIS)

Maarek, V.; Sharir, Y.; Stern, A.

1980-03-01

The effect of electrode vertex angle on penetration depth and weld bead width, in automatic tungsten-inert-gas (TIG) dcsp bead-on-plate welding with different currents, has been studied for stainless steel 304L plates 1.5 mm and 8 mm thick. It has been found that for thin plates, wider and deeper welds are obtained when using sharper electrodes while, for thick plates, narrower and deeper welds are produced when blunt electrodes (vertex angle 180 deg) are used. An explanation of the results, based on a literature survey, is included

Strengthening of Reinforced Concrete Beam in Shear Zone by Compensation the Stirrups with Equivalent External Steel Plates

Directory of Open Access Journals (Sweden)

Khamail Abdul-Mahdi Mosheer

2016-09-01

Full Text Available An experimental study on reinforced concrete beams strengthened with external steel plates instead of shear stirrups has been held in this paper. Eight samples of the same dimensions and properties were used. Two of them were tested up to failure and specified as references beams; one with shear reinforcement and the other without shear reinforcement. Another samples without shear reinforcement were tested until the first shear crack occurs, then the samples strengthened on both sides with external steel plates as equivalent area of removed stirrups. The strengthened beams were divided into three groups according to the thickness of plates (1, 1.5, 2 mm, each group involved two beams; one bonded using epoxy and the other bonded using epoxy with anchored bolts. Finally, the strengthened beams tested when using anchored bolts with epoxy glue to bond plates. Where the increasing in maximum load is higher than that in reference beam with no internal stirrups reach to (75.46 –106.13% and has a good agreement with the control beam with shear reinforcement reach to (76.06 – 89.36% of ultimate load.

Electroplating of Ni-Mo Coating on Stainless Steel for Application in Proton Exchange Membrane Fuel Cell Bipolar Plate

Directory of Open Access Journals (Sweden)

H. Rashtchi

2018-03-01

Full Text Available Stainless steel bipolar plates are preferred choice for use in Proton Exchange Membrane Fuel Cells (PEMFCs. However, regarding the working temperature of 80 °C and corrosive and acidic environment of PEMFC, it is necessary to apply conductive protective coatings resistant to corrosion on metallic bipolar plate surfaces to enhance its chemical stability and performance. In the present study, by applying Ni-Mo and Ni-Mo-P alloy coatings via electroplating technique, corrosion resistance was improved, oxid layers formation on substrates which led to increased electrical conductivity of the surface was reduced and consequently bipolar plates fuction was enhanced. Evaluation tests included microstructural and phase characterizations for evaluating coating components; cyclic voltammetry test for electrochemical behavior investigations; wettability test for measuring hydrophobicity characterizations of the coatings surfaces; interfacial contact resistance measurements of the coatings for evaluating the composition of applied coatings; and polarization tests of fuel cells for evaluating bipolar plates function in working conditions. Finally, the results showed that the above-mentioned coatings considerably decreased the corrosion and electrical resistance of the stainless steel.

The CCT diagrams of ultra low carbon bainitic steels and their impact toughness properties

International Nuclear Information System (INIS)

Lis, A.K.; Lis, J.; Jeziorski, L.

1998-01-01

The CCT diagrams of ULCB N i steels, HN3MV, HN3MVCu having 5.1% Ni and 3.5% Ni and Cu bearing steels; HN3M1.5Cu, HSLA 100 have been determined. The reduced carbon concentration in steel, in order to prevent the formation of cementite, allowed for using nickel, manganese, chromium and molybdenum to enhance hardenability and refinement of the bainitic microstructures by lowering B S temperature. Copper and microadditions of vanadium and niobium are successfully used for precipitation strengthening of steel both in thermomechanically or heat treated conditions. Very good fracture toughness at low temperatures and high yield strength properties of HN3MVCu and HN3MV steels allowed for fulfillment of the requirements for steel plates for pressure vessels and cryogenic applications. (author)

Effect of aging on properties of pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

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

1986-04-01

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

Establishment of quality system program for the manufacture of nuclear grade steels

International Nuclear Information System (INIS)

Saito, Toru

1978-01-01

Recently, the pressure vessels employed in the fields of nuclear power generation, petroleum refining, and chemical industry tend to be large size and high performance, therefore thick steel plates and forgings of high quality are demanded, and the establishment and maintenance of strict quality assurance system for material production are required. In Mizushima Works, Kawasaki Steel Corp., the installations for large forgings and thick plates were constructed from 1969 to 1976, and the researches on the development of production techniques were forwarded in parallel. A Quality System Certificate (Material) has been granted to Mizushima Works on March 11, 1977, by the American Society of Mechanical Engineers for the manufacture of steel plates and forgings of nuclear grade. The quality system manual as the basis of quality assurance activities, the organization for quality assurance, its responsibility and authority, the discrimination of materials, the control of manufacturing processes, the qualification of workers in charge of nondestructive test, welding, heat treatment and others, the management of tests and inspections, the management of products with faults and repairing measures, and internal supervisory system are explained. The wide use of computers is novel in the field of quality system program. (Kako, I.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Ballistic Limit of High-Strength Steel and Al7075-T6 Multi-Layered Plates Under 7.62-mm Armour Piercing Projectile Impact

OpenAIRE

Rahman, N. A.; Abdullah, S.; Zamri, W. F. H.; Abdullah, M. F.; Omar, M. Z.; Sajuri, Z.

2016-01-01

Abstract This paper presents the computational-based ballistic limit of laminated metal panels comprised of high strength steel and aluminium alloy Al7075-T6 plate at different thickness combinations to necessitate the weight reduction of existing armour steel plate. The numerical models of monolithic configuration, double-layered configuration and triple-layered configuration were developed using a commercial explicit finite element code and were impacted by 7.62 mm armour piercing projectil...

JR-curves of wide plates and CT25 specimens

International Nuclear Information System (INIS)

Aurich, D.; Wobst, K.; Krafka, H.

1987-01-01

In connection with the problem of the applicability of the characteristic specimen date - i.e. the initiation and stable crack propagation under maximal loads, together with the elastic-plastic material behaviour - to that of actual components, spot-check type beside tests were conducted using wide-plate central crack, central notch (CCT, CNT) and double external crack (DECT) samples. The material in question was an StE 460 steel. A comparison between the determined values shows that the assessed pressure vessel behaviour differs extensively to the values derived from the CCT and CNT specimens. The corresponding results obtained from the CT25 and DECT specimens vary only slightly in the region of interest and correspond to real vessel values. (orig./DG) [de

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

International Nuclear Information System (INIS)

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

1987-03-01

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

Establishment of welding process without PWHT and preheating in SGV480 plate for nuclear reactor containment vessel

International Nuclear Information System (INIS)

Watanabe, Nozomu; Higashikubo, Tomohiro; Nagamura, Takafumi; Yoshimoto Kentaro

2000-01-01

Ordinances of Japan's Ministry of International Trade and Industry provide that welded joints more than 38 mm thick used in nuclear reactor containment vessels undergo Post Weld Heat Treatment (PWHT). PWHT is difficult to apply in the field, however. We made SGV480 plate tougher and more weldable by using a Thermo-Mechanical Control Process (TMCP) in rolling. Such plate can be used without PWHT or preheating up to 55 mm thick at lowest service temperature -19degC. (author)

Establishment of welding process without PWHT and preheating in SGV480 plate for nuclear reactor containment vessel

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Nozomu; Higashikubo, Tomohiro; Nagamura, Takafumi [Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan); Yoshimoto Kentaro [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan). Takasago Research and Development Center

2000-07-01

Ordinances of Japan's Ministry of International Trade and Industry provide that welded joints more than 38 mm thick used in nuclear reactor containment vessels undergo Post Weld Heat Treatment (PWHT). PWHT is difficult to apply in the field, however. We made SGV480 plate tougher and more weldable by using a Thermo-Mechanical Control Process (TMCP) in rolling. Such plate can be used without PWHT or preheating up to 55 mm thick at lowest service temperature -19degC. (author)

Shallow-crack toughness results for reactor pressure vessel steel

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1974-12-01

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

Reactor pressure vessel steels ASTM A533B and A508 Cl.2

International Nuclear Information System (INIS)

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

1979-11-01

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

Microstructure and mechanical properties of friction stir welded 18Cr–2Mo ferritic stainless steel thick plate

International Nuclear Information System (INIS)

Han, Jian; Li, Huijun; Zhu, Zhixiong; Barbaro, Frank; Jiang, Laizhu; Xu, Haigang; Ma, Li

2014-01-01

Highlights: • We focus on friction stir welding of 18Cr–2Mo ferritic stainless steel thick plate. • We produce high-quality joints with special tool and optimised welding parameters. • We compare microstructure and mechanical properties of steel and joint. • Friction stir welding is a method that can maintain the properties of joint. - Abstract: In this study, microstructure and mechanical properties of a friction stir welded 18Cr–2Mo ferritic stainless steel thick plate were investigated. The 5.4 mm thick plates with excellent properties were welded at a constant rotational speed and a changeable welding speed using a composite tool featuring a chosen volume fraction of cubic boron nitride (cBN) in a W–Re matrix. The high-quality welds were successfully produced with optimised welding parameters, and studied by means of optical microscopy (OM), scanning electron microscopy (SEM), electron back-scattered diffraction (EBSD) and standard hardness and impact toughness testing. The results show that microstructure and mechanical properties of the joints are affected greatly, which is mainly related to the remarkably fine-grained microstructure of equiaxed ferrite that is observed in the friction stir welded joint. Meanwhile, the ratios of low-angle grain boundary in the stir zone regions significantly increase, and the texture turns strong. Compared with the base material, mechanical properties of the joint are maintained in a comparatively high level

Characterization of matrix damage in ion-irradiated reactor vessel steel

International Nuclear Information System (INIS)

Fujii, Katsuhiko; Fukuya, Koji

2004-01-01

Exact nature of the matrix damage, that is one of radiation-induced nano-scale microstructural features causing radiation embrittlement of reactor vessel, in irradiated commercial steels has not been clarified yet by direct characterization using transmission electron microscopy (TEM). We designed a new preparation method of TEM observation samples and applied it to the direct TEM observation of the matrix damage in the commercial steel samples irradiated by ions. The simulation irradiation was carried out by 3 MeV Ni 2+ ion to a dose of 1 dpa at 290degC. Thin foil specimens for TEM observation were prepared using the modified focused ion beam method. A weak-beam TEM study was carried out for the observation of matrix damage in the samples. Results of this first detailed observation of the matrix damage in the irradiated commercial steel show that it is consisted of small dislocation loops. The observed and analyzed dislocation loops have Burgers vectors b = a , and a mean image size and the number density are 2.5 nm and about 1 x 10 22 m -3 , respectively. In this experiment, all of the observed dislocation loops were too small to determine the vacancy or interstitial nature of the dislocation loops directly. Although it is an indirect method, post-irradiation annealing was used to infer the loop nature. Most of dislocation loops were stable after the annealing at 400degC for 30 min. This result suggests that their nature is interstitial. (author)

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Studies on the temperature distribution of steel plates with different paints under solar radiation

International Nuclear Information System (INIS)

Liu, Hongbo; Chen, Zhihua; Chen, Binbin; Xiao, Xiao; Wang, Xiaodun

2014-01-01

Thermal effects on steel structures exposed to solar radiation are significant and complicated. Furthermore, the solar radiation absorption coefficient of steel surface with different paintings is the main factor affecting the non-uniform temperature of spatial structures under solar radiation. In this paper, nearly two hundreds steel specimens with different paintings were designed and measured to obtain their solar radiation absorption coefficients using spectrophotometer. Based on the test results, the effect of surface color, painting type, painting thickness on the solar radiation absorption coefficient was analyzed. The actual temperatures under solar radiation for all specimens were also measured in summer not only to verify the absorption coefficient but also provide insight for the temperature distribution of steel structures with different paintings. A numerical simulation and simplified formula were also conducted and verified by test, in order to study the temperature distribution of steel plates with different paints under solar radiation. The results have given an important reference in the future research of thermal effect of steel structures exposed to solar radiation. - Highlights: • Solar radiation absorptions for steel with different paintings were measured. • The temperatures of all specimens under solar radiation were measured. • The effect of color, thickness and painting type on solar absorption was analyzed. • A numerical analysis was conducted and verified by test data. • A simplified formula was deduced and verified by test data

Ricochet of a tungsten heavy alloy long-rod projectile from deformable steel plates

International Nuclear Information System (INIS)

Lee, Woong; Lee, Heon-Joo; Shin, Hyunho

2002-01-01

Ricochet of a tungsten heavy alloy long-rod projectile from oblique steel plates with a finite thickness was investigated numerically using a full three-dimensional explicit finite element method. Three distinctive regimes resulting from oblique impact depending on the obliquity, namely simple ricochet, critical ricochet and target perforation, were investigated in detail. Critical ricochet angles were calculated for various impact velocities and strengths of the target plates. It was predicted that critical ricochet angle increases with decreasing impact velocities and that higher ricochet angles were expected if higher strength target materials are employed. Numerical predictions were compared with existing two-dimensional analytical models. Experiments were also carried out and the results supported the predictions of the numerical analysis

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

International Nuclear Information System (INIS)

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

1993-12-01

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

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

1993-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

Microstructures and Mechanical Properties of Austempering SUS440 Steel Thin Plates

Directory of Open Access Journals (Sweden)

Cheng-Yi Chen

2016-02-01

Full Text Available SUS440 is a high-carbon stainless steel, and its martensite matrix has high heat resistance, high corrosion resistance, and high pressure resistance. It has been widely used in mechanical parts and critical materials. However, the SUS440 martempered matrix has reliability problems in thin plate applications and thus research uses different austempering heat treatments (tempering temperature: 200 °C–400 °C to obtain a matrix containing bainite, retained austenite, martensite, and the M7C3 phase to investigate the relationships between the resulting microstructure and tensile mechanical properties. Experimental data showed that the austempering conditions of the specimen affected the volume fraction of phases and distribution of carbides. After austenitizing heat treatment (1080 °C for 30 min, the austempering of the SUS440 thin plates was carried out at a salt-bath temperature 300 °C for 120 min and water quenching was then used to obtain the bainite matrix with fine carbides, with the resulting material having a higher tensile fracture strength and average hardness (HRA 76 makes it suitable for use as a high-strength thin plate for industrial applications.

Corrosion resistance of a magnetic stainless steel ion-plated with titanium nitride.

Science.gov (United States)

Hai, K; Sawase, T; Matsumura, H; Atsuta, M; Baba, K; Hatada, R

2000-04-01

This in vitro study evaluated the corrosion resistance of a titanium nitride (TiN) ion-plated magnetic stainless steel (447J1) for the purpose of applying a magnetic attachment system to implant-supported prostheses made of titanium. The surface hardness of the TiN ion-plated 447J1 alloy with varying TiN thickness was determined prior to the corrosion testing, and 2 micrometers thickness was confirmed to be appropriate. Ions released from the 447J1 alloy, TiN ion-plated 447J1 alloy, and titanium into a 2% lactic acid aqueous solution and 0.1 mol/L phosphate buffered saline (PBS) were determined by means of an inductively coupled plasma atomic emission spectroscopy (ICP-AES). Long-term corrosion behaviour was evaluated using a multisweep cyclic voltammetry. The ICP-AES results revealed that the 447J1 alloy released ferric ions into both media, and that the amount of released ions increased when the alloy was coupled with titanium. Although both titanium and the TiN-plated 447J1 alloy released titanium ions into lactic acid solution, ferric and chromium ions were not released from the alloy specimen for all conditions. Cyclic voltamograms indicated that the long-term corrosion resistance of the 447J1 alloy was considerably improved by ion-plating with TiN.

Residual strains in a stainless steel perforated plate subjected to reverse loading at high temperature

International Nuclear Information System (INIS)

Durelli, A.J.; Buitrago, J.

1974-01-01

An investigation was made to determine strains in a stainless steel perforated plate subjected to a temperature of 1100 0 F and to a successively applied tensile and compressive in-plane loading sufficiently large to produce creep and plastic strains. The duration of the test was 1000 hours. Square grids of lines (at distance of 0.25 in.) and crossed-gratings (500 lines-per-inch) were engraved on both surfaces of the plate before the test. After the plate was unloaded and brought back to room temperature the grids were analyzed using traveling microscopes, and the gratings using the moire effect. Both Cartesian strains were determined from the moire isothetics along the axes of the plate, along the two lines tangent to the hole and parallel to those axes and along the edges of the plate. Grid measurements were made at specific points. The deformed shapes of the hole and of the plate are also given. It is estimated that strains larger than 0.001 can be determined with the techniques and methods used. (U.S.)

Damage sensing and mechanical characteristics of CFRP strengthened steel plate

Science.gov (United States)

Mieda, Genki; Nakano, Daiki; Fuji, Yuya; Nakamura, Hitoshi; Mizuno, Yosuke; Nakamura, Kentaro; Matsui, Takahiro; Ochi, Yutaka; Matsumoto, Yukihiro

2017-10-01

In recent years, a large number of structures that were built during the period of high economic growth in Japan is beginning to show signs of aging. For example, the structural performance of steel structures has degraded due to corrosion. One measure that has been proposed and studied to address this issue is the adhesive bonding method, which can be used to repair and reinforce these structures. However, this method produces brittle fracture in the adhesive layer and is difficult to maintain after bonding. To solve the problem faced by this method, a clarification of the mechanical properties inside the adhesive is necessary. Then this background, a fiber Bragg grating (FBG) sensor has been used in this study. This sensor can be embedded within the building material that needs repairing and reinforcing because an FBG sensor is extremely small. Eventually based on this, a three-point bending test of a carbon fiber reinforced plastic (CFRP) strengthened steel plate that was embedded with an FBG sensor was conducted. This paper demonstrates that an FBG sensor is effectively applicable for sensing when damage occurs.

Stable and unstable crack growth in Type 304 stainless steel plate

International Nuclear Information System (INIS)

Yagawa, G.

1984-01-01

Experimental and theoretical results on stable as well as unstable fractures for Type 304 stainless steel plates with a central crack subjected to tension force are given. In the experiment using a testing machine with a special spring for high compliance, the transition points from the stable to the unstable crack growth are observed and comparisons are made between the test results and the finite element solutions. A round robin calculation for the elastic-plastic stable crack growth using one of the specimens mentioned above is also given. (orig.)

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

International Nuclear Information System (INIS)

Whitman, G.D.

1986-03-01

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

Prestressed pressure vessel for nuclear power plants

International Nuclear Information System (INIS)

1974-01-01

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

Hydrogen effect on mechanical properties and flake formation in the 10KhSND steel rolled plates

Energy Technology Data Exchange (ETDEWEB)

Muradova, R G; Zakharov, V A; Kuzin, A P; Gol' tsov, V A; Podgajskij, M S [Donetskij Politekhnicheskij Inst. (Ukrainian SSR); Donetskij Nauchno-Issledovatel' skij Inst. Chernoj Metallurgii (Ukrainian SSR))

1982-01-01

The effect of hydrogen on mechanical properties of the 10KhSND steel rolled plates during natural aging is studied. Optimum period of metal acceptance tests, which are advisable to conduct after 5-7 day natural aging of finished products, are found out. The technique is worked out and a safe hydrogen content to prevent flake formation in the 10KhSND steel is determined. It is shown that a safe hydrogen content is dependent on the experiment conditions (sample dimensions, conditions of cooling, and prehistory).

Hydrogen effect on mechanical properties and flake formation in the 10KhSND steel rolled plates

International Nuclear Information System (INIS)

Muradova, R.G.; Zakharov, V.A.; Kuzin, A.P.; Gol'tsov, V.A.; Podgajskij, M.S.

1982-01-01

The effect of hydrogen on mechanical properties of the 10KhSND steel rolled plates during natural aging is studied. Optimum period of metal acceptance tests, which are advisable to conduct after 5-7 day natural aging of finished products, are found out. The technique is worked out and a safe hydrogen content to prevent flake formation in the 10KhSND steel is determined. It is shown that a safe hydrogen content is dependent on the experiment conditions (sample dimensions, conditions of cooling, and prehistory)

Design and analysis of reactor containment of steel-concrete composite laminated shell

International Nuclear Information System (INIS)

Ichikawa, K.; Isobata, O.; Kawamata, S.

1977-01-01

A new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. Because of the compressive and shearing resistance of the concrete core, the layers behave as a composite solid shell. Membrane forces are shared by steel plates and partly by concrete core. Bending moment is effectively resisted by the section with extreme layers of steel. Therefore, both surfaces can be designed as extremely thin plates: the inner plate, which is a load carrying members as well as a liner, can be welded without the laborious process of stress-relieving, and various jointing methods can be applied to the outer plate which is free from the need for leak tightness. The capability of the composite layers of behaving as a unified solid shell section depends largely on the shearing rigidity of the concrete core. However, as its resisting capacity to transverse shearing force is comparatively low, a device for reducing the shearing stress at the junction to the base mat is needed. In the new scheme, this part of the cylindrical shell is divided into multiple layers of the same kind of composite shell. This device makes the stiffness of the bottom of the cylindrical shell to lateral movement minimum while maintaining the proper resistance to membrane forces. The analysis shows that the transverse shearing stress can be reduced to less than 1√n of the ordinary case by dividing the thickness of the shell into n layers which are able to slip against each other at the contact surface. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented

Reduction of core loss in non-oriented (NO) electrical steel by electroless-plated magnetic coating

International Nuclear Information System (INIS)

Chivavibul, Pornthep; Enoki, Manabu; Konda, Shigeru; Inada, Yasushi; Tomizawa, Tamotsu; Toda, Akira

2011-01-01

An important issue in development of electrical steels for core-laminated products is to reduce core loss to improve energy conversion efficiency. This is usually obtained by tailoring the composition, microstructure, and texture of electrical steels themselves. A new technique to reduce core loss in electrical steel has been investigated. This technique involves electroless plating of magnetic thin coating onto the surface of electrical steel. The material system was electroless Ni-Co-P coatings with different thicknesses (1, 5, and 10 μm) deposited onto the surface of commercially available Fe-3% Si electrical steel. Characterization of deposited Ni-Co-P coating was carried out using X-ray diffraction (XRD), scanning electron microscope (SEM), and energy dispersive X-ray (EDX) spectrometer. The deposited Ni-Co-P coatings were amorphous and composed of 56-59% Ni, 32-35% Co, and 8-10% P by mass. The effect of coatings on core loss of the electrical steel was determined using single sheet test. A core loss reduction of 4% maximum was achieved with the Ni-Co-P coating of 1 μm thickness at 400 Hz and 0.3 T. - Research Highlights: → New approach to reduce core loss of electrical steel by magnetic coating. → Ni-Co-P coating influences core loss of NO electrical steel. → Core loss increases in RD direction but reduces in TD direction.

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

International Nuclear Information System (INIS)

Ishino, Shiori

2007-01-01

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

Chromate coating of zinc-aluminum plating on mild steel

International Nuclear Information System (INIS)

Haque, I.; Khan, A.; Nadeem, A.

2005-01-01

The chromate coating on zinc-aluminium deposits has been studied. Zinc-aluminium deposition from non-cyanide bath was carried out at current density 3-3.5 A/dm/sup 2/, plating voltage approx. equal to 1.25 V, temperature 18-20 deg. C, for 15 min. The effect of aluminium chloride on the rest potentials of golden, colorless and non-chromated zinc-aluminium alloy deposits was observed. It was found that rest potential was slightly increased with the increase in the concentration of aluminium chloride, only in the case of golden chromating. The rest potential of colorless chromated zinc-aluminium deposits on mild steel were observed to have no correlation with aluminium chloride concentration. The abrasion resistance of colorless chromating was better than golden chromating. (author)

Investigation on impact resistance of steel plate reinforced concrete barriers against aircraft impact. Pt.3: Analyses of full-scale aircraft impact

International Nuclear Information System (INIS)

Jun Mizuno; Norihide Koshika; Eiichi Tanaka; Atsushi Suzuki; Yoshinori Mihara; Isao Nishimura

2005-01-01

Steel plate reinforced concrete (SC) walls and slabs are structural members in which the rebars of reinforced concrete are replaced by steel plates. Steel plate reinforced concrete structures are more attractive structural design alternatives to reinforced concrete structures, especially with thick, heavily reinforced walls and slabs such as nuclear structures, because they enable a much shorter construction period, greater earthquake resistant and more cost effectiveness. Experimental and analytical studies performed by the authors have also shown that SC structures are much more effective in mitigating damage against scaled aircraft models , as described in Parts 1 and 2 of this study. The objective of Part 3 was to determine the protective capability of SC walls and roofs against a full-scale aircraft impact by conducting numerical experiments to investigate the fracture behaviors and limit thicknesses of SC panels and to examine the effectiveness of SC panels in detail under design conditions. Furthermore, a simplified method is proposed for evaluating the localized damage induced by a full-scale engine impact. (authors)

76 FR 54207 - Stainless Steel Plate in Coils From Italy: Revocation of Antidumping Duty Order

Science.gov (United States)

2011-08-31

... continuation or recurrence of material injury to an industry in the United States within a reasonably foreseeable time. See Stainless Steel Plate From Belgium, Italy, Korea, South Africa, and Taiwan, 76 FR 50495..., Italy, Korea, South Africa, and Taiwan pursuant to section 751(c) of the Act. See Initiation. On July 20...

Stress corrosion cracking of nuclear reactor pressure vessel and piping steels

International Nuclear Information System (INIS)

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

1988-01-01

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

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

International Nuclear Information System (INIS)

Voigt, J.

1978-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

Steel plates and concrete filled composite shear walls related nuclear structural engineering: Experimental study for out-of-plane cyclic loading

International Nuclear Information System (INIS)

Li, Xiaohu; Li, Xiaojun

2017-01-01

Based on the program of CAP1400 nuclear structural engineering, the out-of-plane seismic behavior of steel plate and concrete infill composite shear walls (SCW) was investigated. 6 1/5 scaled specimens were conducted which consist of 5 SCW specimens and 1 reinforced concrete (RC) specimen. The specimens were tested under out-of-plane cyclic loading. The effect of the thickness of steel plate, vertical load and the strength grade of concrete on the out-of-plane seismic behavior of SCW were analyzed. The results show that the thickness of steel plate and vertical load have great influence on the ultimate bearing capacity and lateral stiffness, however, the influence of the strength grade of concrete was little within a certain range. SCW is presented to have a better ultimate capacity and lateral stiffness but have worse ductility in failure stage than that of RC. Based on the experiment, the cracking load of concrete infill SCW was analyzed in theory. The modified calculation formula of the cracking load was made, the calculated results showed good agreement with the test results. The formula can be used as the practical design for the design of cracking loads.

Steel plates and concrete filled composite shear walls related nuclear structural engineering: Experimental study for out-of-plane cyclic loading

Energy Technology Data Exchange (ETDEWEB)

Li, Xiaohu [The College of Architecture and Civil Engineering, Beijing University of Technology, Beijing 100124 (China); Li, Xiaojun, E-mail: beerli@vip.sina.com [The College of Architecture and Civil Engineering, Beijing University of Technology, Beijing 100124 (China); Institute of Geophysics, China Earthquake Administration, Beijing 100081 (China)

2017-04-15

Based on the program of CAP1400 nuclear structural engineering, the out-of-plane seismic behavior of steel plate and concrete infill composite shear walls (SCW) was investigated. 6 1/5 scaled specimens were conducted which consist of 5 SCW specimens and 1 reinforced concrete (RC) specimen. The specimens were tested under out-of-plane cyclic loading. The effect of the thickness of steel plate, vertical load and the strength grade of concrete on the out-of-plane seismic behavior of SCW were analyzed. The results show that the thickness of steel plate and vertical load have great influence on the ultimate bearing capacity and lateral stiffness, however, the influence of the strength grade of concrete was little within a certain range. SCW is presented to have a better ultimate capacity and lateral stiffness but have worse ductility in failure stage than that of RC. Based on the experiment, the cracking load of concrete infill SCW was analyzed in theory. The modified calculation formula of the cracking load was made, the calculated results showed good agreement with the test results. The formula can be used as the practical design for the design of cracking loads.

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Tensile Stress-Strain Results for 304L and 316L Stainless-Steel Plate at Temperature

International Nuclear Information System (INIS)

R. K. Blandford; D. K. Morton; S. D. Snow; T. E. Rahl

2007-01-01

The Idaho National Laboratory (INL) is conducting moderate strain rate (10 to 200 per second) research on stainless steel materials in support of the Department of Energy's (DOE) National Spent Nuclear Fuel Program (NSNFP). For this research, strain rate effects are characterized by comparison to quasi-static tensile test results. Considerable tensile testing has been conducted resulting in the generation of a large amount of basic material data expressed as engineering and true stress-strain curves. The purpose of this paper is to present the results of quasi-static tensile testing of 304/304L and 316/316L stainless steels in order to add to the existing data pool for these materials and make the data more readily available to other researchers, engineers, and interested parties. Standard tensile testing of round specimens in accordance with ASTM procedure A 370-03a were conducted on 304L and 316L stainless-steel plate materials at temperatures ranging from -20 F to 600 F. Two plate thicknesses, eight material heats, and both base and weld metal were tested. Material yield strength, Young's modulus, ultimate strength, ultimate strain, failure strength and failure strain were determined, engineering and true stress-strain curves to failure were developed, and comparisons to ASME Code minimums were made. The procedures used during testing and the typical results obtained are described in this paper

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...

Vibration-response due to thickness loss on steel plate excited by resonance frequency

Science.gov (United States)

Kudus, S. A.; Suzuki, Y.; Matsumura, M.; Sugiura, K.

2018-04-01

The degradation of steel structure due to corrosion is a common problem found especially in the marine structure due to exposure to the harsh marine environment. In order to ensure safety and reliability of marine structure, the damage assessment is an indispensable prerequisite for plan of remedial action on damaged structure. The main goal of this paper is to discuss simple vibration measurement on plated structure to give image on overview condition of the monitored structure. The changes of vibration response when damage was introduced in the plate structure were investigated. The damage on plate was simulated in finite element method as loss of thickness section. The size of damage and depth of loss of thickness were varied for different damage cases. The plate was excited with lower order of resonance frequency in accordance estimate the average remaining thickness based on displacement response obtain in the dynamic analysis. Significant reduction of natural frequency and increasing amplitude of vibration can be observed in the presence of severe damage. The vibration analysis summarized in this study can serve as benchmark and reference for researcher and design engineer.

Super-low-frequency wireless power transfer with lightweight coils for passing through a stainless steel plate

Science.gov (United States)

Ishida, Hiroki; Kyoden, Tomoaki; Furukawa, Hiroto

2018-03-01

To achieve wireless power transfer (WPT) through a stainless-steel plate, a super-low frequency (SLF) was used as a resonance frequency. In our previous study of SLF-WPT, heavy coils were prepared. In this study, we designed lightweight coils using a WPT simulator that we developed previously. As a result, the weight was reduced to 1.69 kg from 11.9 kg, the previous coil weight. At a resonance frequency of 400 Hz, the transmission efficiency and output power of advanced SLF-WPT reached 91% and 426 W, respectively, over a transmission distance of 30 mm. Furthermore, 80% efficiency and 317 W output were achieved when transmitting power through a 1 mm-thick stainless-steel plate. This performance is much better than that in previous reports. We show using both calculations and experimental results that a power-to-weight ratio of 252 W/kg is possible even when using a 400 Hz power supply frequency.

Modeling Fragment Simulating Projectile Penetration into Steel Plates Using Finite Elements and Meshfree Particles

Directory of Open Access Journals (Sweden)

James O’Daniel

2011-01-01

Full Text Available Simulating fragment penetration into steel involves complicated modeling of severe behavior of the materials through multiple phases of response. Penetration of a fragment-like projectile was simulated using finite element (FE and meshfree particle formulations. Extreme deformation and failure of the material during the penetration event were modeled with several approaches to evaluate each as to how well it represents the actual physics of the material and structural response. A steel Fragment Simulating Projectile (FSP – designed to simulate a fragment of metal from a weapon casing – was simulated for normal impact into a flat square plate. A range of impact velocities was used to examine levels of exit velocity ranging from relatively small to one on the same level as the impact velocity. The numerical code EPIC, used for all the simulations presented herein, contains the element and particle formulations, as well as the explicit methodology and constitutive models needed to perform these simulations. These simulations were compared against experimental data, evaluating the damage caused to the projectile and the target plates, as well as comparing the residual velocity when the projectile perforated the target.

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

1994-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

Welding of Thin Steel Plates by Hybrid Welding Process Combined TIG Arc with YAG Laser

Science.gov (United States)

Kim, Taewon; Suga, Yasuo; Koike, Takashi

TIG arc welding and laser welding are used widely in the world. However, these welding processes have some advantages and problems respectively. In order to improve problems and make use of advantages of the arc welding and the laser welding processes, hybrid welding process combined the TIG arc with the YAG laser was studied. Especially, the suitable welding conditions for thin steel plate welding were investigated to obtain sound weld with beautiful surface and back beads but without weld defects. As a result, it was confirmed that the shot position of the laser beam is very important to obtain sound welds in hybrid welding. Therefore, a new intelligent system to monitor the welding area using vision sensor is constructed. Furthermore, control system to shot the laser beam to a selected position in molten pool, which is formed by TIG arc, is constructed. As a result of welding experiments using these systems, it is confirmed that the hybrid welding process and the control system are effective on the stable welding of thin stainless steel plates.

Effects of multi-pass arc welding on mechanical properties of carbon steel C25 plate

International Nuclear Information System (INIS)

Adedayo, S.M.; Babatunde, A.S.

2013-01-01

The effects of multi-pass welding on mechanical properties of C25 carbon steel plate were examined. Mild steel plate workpieces of 90 x 55 mm 2 area and 10 mm thickness with a 30 degrees vee weld-grooves were subjected to single and multi-pass welding. Toughness, hardness and tensile tests of single and multi-pass welds were conducted. Toughness values of the welds under double pass welds were higher than both single pass and unwelded alloy, at respective maximum values of 2464, 2342 and 2170 kN/m. Hardness values were reduced under double pass relative to single pass welding with both being lower than the value for unwelded alloy; the values were 40.5, 43.2 and 48.5 Rs respectively at 12 mm from the weld line. The tensile strength of 347 N/mm 2 under multi-pass weld was higher than single pass weld with value of 314 N/mm 2 . Therefore, the temperature distribution and apparent pre-heating during multi-pass welding increased the toughness and tensile strength of the weldments, but reduced the hardness. (au)

Gigacycle fatigue behaviour of austenitic stainless steels used for mercury target vessels

International Nuclear Information System (INIS)

Naoe, Takashi; Xiong, Zhihong; Futakawa, Masatoshi

2016-01-01

A mercury enclosure vessel for the pulsed spallation neutron source manufactured from a type 316L austenitic stainless steel, a so-called target vessel, suffers the cyclic loading caused by the proton beam induced pressure waves. A design criteria of the JSNS target vessel which is defined based on the irradiation damage is 2500 h at 1 MW with a repetition rate of 25 Hz, that is, the target vessel suffers approximately 10 9 cyclic loading while in operation. Furthermore, strain rate of the beam window of the target vessel reaches 50 s −1 at the maximum, which is much higher than that of the conventional fatigue. Gigacycle fatigue strength up to 10 9 cycles for solution annealed 316L (SA) and cold-worked 316L (CW) were investigated through the ultrasonic fatigue tests. Fatigue tests were performed under room temperature and 250 °C which is the maximum temperature evaluated at the beam window in order to investigate the effect of temperature on fatigue strength of SA and CW 316L. The results showed that the fatigue strength at 250 °C is clearly reduced in comparison with room temperature, regardless of cold work level. In addition, residual strength and microhardness of the fatigue tested specimen were measured to investigate the change in mechanical properties by cyclic loading. Cyclic hardening was observed in both the SA and CW 316L, and cyclic softening was observed in the initial stage of cyclic loading in CW 316L. Furthermore, abrupt temperature rising just before fatigue failure was observed regardless of testing conditions.

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

International Nuclear Information System (INIS)

Miannay, D.

1980-05-01

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

76 FR 53882 - Continuation of Antidumping and Countervailing Duty Orders: Stainless Steel Plate in Coils From...

Science.gov (United States)

2011-08-30

... Coils From Belgium, the Republic of Korea, South Africa, and Taiwan AGENCY: Import Administration... steel plate in coils (SSPC) from Belgium, the Republic of Korea (Korea), South Africa, and Taiwan would... and CVD orders would likely lead to a continuation or recurrence of material injury to an industry in...

Weld repair of helium degraded reactor vessel material

International Nuclear Information System (INIS)

Kanne, W.R. Jr.; Lohmeier, D.A.; Louthan, M.R. Jr.; Rankin, D.T.; Franco-Ferreira, E.A.; Bruck, G.J.; Madeyski, A.; Shogan, R.P.; Lessmann, G.G.

1990-01-01

Welding methods for modification or repair of irradiated nuclear reactor vessels are being evaluated at the Savannah River Site. A low-penetration weld overlay technique has been developed to minimize the adverse effects of irradiation induced helium on the weldability of metals and alloys. This technique was successfully applied to Type 304 stainless steel test plates that contained 3 to 220 appm helium from tritium decay. Conventional welding practices caused significant cracking and degradation in the test plates. Optical microscopy of weld surfaces and cross sections showed that large surface toe cracks formed around conventional welds in the test plates but did not form around overlay welds. Scattered incipient underbead cracks (grain boundary separations) were associated with both conventional and overlay test welds. Tensile and bend tests were used to assess the effect of base metal helium content on the mechanical integrity of the low-penetration overlay welds. The axis of tensile specimens was perpendicular to the weld-base metal interface. Tensile specimens were machined after studs were resistance welded to overlay surfaces

Radioactive waste processing vessel

International Nuclear Information System (INIS)

Hayashi, Masaru; Suzuki, Osamu; Ishizaki, Kanjiro.

1987-01-01

Purpose: To obtain a vessel of a reduced weight and with no external leaching of radioactive materials. Constitution: The vessel main body is constituted, for example, with light weight concretes or foamed concretes, particularly, foamed concretes containing fine closed bubbles in the inside. Then, layers having dense texture made of synthetic resin such as polystylene, vinylchloride resin, etc. or metal plate such as stainless plate are integrally disposed to the inner surface of the vessel main body. The cover member also has the same structure. (Sekiya, K.)

Survey of postirradiation heat treatment as a means to mitigate radiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Hawthorne, J.R.

1979-01-01

Nuclear-radiation service typically produces a progressive reduction in the notch ductility of low-alloy steels. The reduction is manifested by a decrease in Charpy-V (Csub(v)) upper-shelf energy level and by an elevation in temperature of the ductile-to-brittle transition. Post irradiation heat treatment (annealing) is being investigated as a method for the reversal of these detrimental radiation effects for reactor-vessel steels. This study was undertaken to analyze factors which could affect annealing response, report data available to qualify suspected influences on annealing, and summarize experimental results generated for many commercially produced reactor materials and companion materials produced in the laboratory

Steel impurity element effects on postirradiation properties recovery by annealing

International Nuclear Information System (INIS)

Hawthorne, J.R.

1987-01-01

The influence of copper content and phosphorus content on notch ductility recovery by 399 0 C postirradiation heat treatment was explored for A 533-B and A 302-B pressure vessel steels. Charpy-V (C/sub V/) specimens for the investigation were obtained from ten plates produced from four (4-way split) laboratory melts. The plates were 15.2 mm thick but were heat treated to reproduce the microstructure of 150-mm and thicker A 302-B plates at the quarter-thickness location. The C/sub V/ specimens were irradiated in two assemblies at 288 0 C (550 0 F) to a fluence of --2.5 x 10/sup 19/ n/cm/sup 2/ in a light-water-cooled and moderated test reactor. Notch ductility properties in the asirradiated and 399 0 C, 168 h postirradiation-annealed conditions were determined. In addition, separate sets of specimens were thermally conditioned at 288 0 C and at 288 0 C followed by 399 0 C to benchmark the effects of temperature in the absence of irradiation. The results indicate that copper has a significant influence on the magnitude of residual embrittlement after annealing. In contrast, phosphorus contents in the range of 0.002 to 0.025% were found not to have an effect on residual embrittlement either in high or low copper steels. Essentially full recovery in 41-J transition temperature was observed for high phosphorus, low copper content steels. Effects of nickel alloying on recovery behavior were also investigated through data comparisons for A 302-B versus A 533-B plates

Optimizing the Steel Plate Storage Yard Crane Scheduling Problem Using a Two Stage Planning/Scheduling Approach

DEFF Research Database (Denmark)

Hansen, Anders Dohn; Clausen, Jens

This paper presents the Steel Plate Storage Yard Crane Scheduling Problem. The task is to generate a schedule for two gantry cranes sharing tracks. The schedule must comply with a number of constraints and at the same time be cost efficient. We propose some ideas for a two stage planning...

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-12-31

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

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.

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)

Fracture toughness behavior and its analysis on nuclear pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

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

1983-02-01

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

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

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

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