Helium in inert matrix dispersion fuels

International Nuclear Information System (INIS)

Veen, A. van; Konings, R.J.M.; Fedorov, A.V.

2003-01-01

The behaviour of helium, an important decay product in the transmutation chains of actinides, in dispersion-type inert matrix fuels is discussed. A phenomenological description of its accumulation and release in CERCER and CERMET fuel is given. A summary of recent He-implantation studies with inert matrix metal oxides (ZrO 2 , MgAl 2 O 4 , MgO and Al 2 O 3 ) is presented. A general picture is that for high helium concentrations helium and vacancy defects form helium clusters which convert into over-pressurized bubbles. At elevated temperature helium is released from the bubbles. On some occasions thermal stable nano-cavities or nano-pores remain. On the basis of these results the consequences for helium induced swelling and helium storage in oxide matrices kept at 800-1000 deg. C will be discussed. In addition, results of He-implantation studies for metal matrices (W, Mo, Nb and V alloys) will be presented. Introduction of helium in metals at elevated temperatures leads to clustering of helium to bubbles. When operational temperatures are higher than 0.5 melting temperature, swelling and helium embrittlement might occur

Effects of helium implantation on fatigue properties of F82H-IEA heat

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, N.; Murase, Y.; Nagakawa, J. [National Research Institute for Metals, Tsukuba, Ibaraki (Japan)

2007-07-01

Full text of publication follows: Ferritic steels including reduced activation ones that have been recognized as attractive structural candidates for DEMO reactors and the beyond are known to be highly resistant to helium embrittlement. However, almost studies that deduced this behavior have been carried out by means of short time experiments such as tensile tests, and a few results are available concerning long term inspections, although the detrimental helium effect appears more severely in the latter. The aim of this work is to obtain further information on the influence of helium on fatigue properties of a representative reduced activation ferritic/martensitic steel F82H (8Cr2WVTa) using helium implantation technique with a cyclotron. The material examined is an IEA heat version of F82H. In order to realize a fine grain size due to thin specimens (0.08 mm thick) for ion irradiation, normalizing was conducted at rather low temperature of 1213 K, followed by tempering at 1023 K. Helium was implanted by {alpha}-particle irradiation at 823 K, a desired highest temperature of this material for first wall application, to the concentration of 100 appm He with an implantation rate of about 1.7 x 10{sup -3} appm He/s. Subsequent fatigue tests were conducted at the same temperature as that of irradiation, not only on implanted specimens but also on reference controls which were not implanted with helium but experienced the same metallurgical histories as those of irradiated ones. After fracture, samples were observed with electron microscopes. In short time periods, it has been notified that helium introduction caused no significant deterioration of both fatigue life and extension at fracture. In addition, all specimens failed in a fully trans-crystalline and ductile manner, irrespective of whether helium was present or not. Indication of grain boundary embrittlement was therefore not discerned. These facts would reflect insusceptible characteristics of this material to

Effects of helium implantation on fatigue properties of F82H-IEA heat

International Nuclear Information System (INIS)

Yamamoto, N.; Murase, Y.; Nagakawa, J.

2007-01-01

Full text of publication follows: Ferritic steels including reduced activation ones that have been recognized as attractive structural candidates for DEMO reactors and the beyond are known to be highly resistant to helium embrittlement. However, almost studies that deduced this behavior have been carried out by means of short time experiments such as tensile tests, and a few results are available concerning long term inspections, although the detrimental helium effect appears more severely in the latter. The aim of this work is to obtain further information on the influence of helium on fatigue properties of a representative reduced activation ferritic/martensitic steel F82H (8Cr2WVTa) using helium implantation technique with a cyclotron. The material examined is an IEA heat version of F82H. In order to realize a fine grain size due to thin specimens (0.08 mm thick) for ion irradiation, normalizing was conducted at rather low temperature of 1213 K, followed by tempering at 1023 K. Helium was implanted by α-particle irradiation at 823 K, a desired highest temperature of this material for first wall application, to the concentration of 100 appm He with an implantation rate of about 1.7 x 10 -3 appm He/s. Subsequent fatigue tests were conducted at the same temperature as that of irradiation, not only on implanted specimens but also on reference controls which were not implanted with helium but experienced the same metallurgical histories as those of irradiated ones. After fracture, samples were observed with electron microscopes. In short time periods, it has been notified that helium introduction caused no significant deterioration of both fatigue life and extension at fracture. In addition, all specimens failed in a fully trans-crystalline and ductile manner, irrespective of whether helium was present or not. Indication of grain boundary embrittlement was therefore not discerned. These facts would reflect insusceptible characteristics of this material to high

Control of helium effects in irradiated materials based on theory and experiment

International Nuclear Information System (INIS)

Mansur, L.K.; Lee, E.H.; Maziasz, P.J.; Rowcliffe, A.F.

1986-01-01

Helium produced in materials by (n,α) transmutation reactions during neutron irradiations or subjected in ion bombardment experiments causes substantial changes in the response to displacement damage. In particular, swelling, phase transformations and embrittlement are strongly affected. Present understanding of the mechanisms underlying these effects is reviewed. Key theoretical relationships describing helium effects on swelling and helium diffusion are described. Experimental data in the areas of helium effects on swelling and precipitation is reviewed with emphasis on critical experiments that have been designed and evaluated in conjunction with theory. Confirmed principles for alloy design to control irradiation performance are described

Molecular dynamics study of the role of symmetric tilt grain boundaries on the helium distribution in nickel

Science.gov (United States)

Torres, E.; Pencer, J.

2018-04-01

Helium impurities, from either direct implantation or transmutation reactions, have been associated with embrittlement in nickel-based alloys. Helium has very low solubility in nickel, and has been found to aggregate at lattice defects such as vacancies, dislocations, and grain boundaries. The retention and precipitation of helium in nickel-based alloys have deleterious effects on the material mechanical properties. However, the underlying mechanisms that lead to helium effects in the host metal are not fully understood. In the present work, we investigate the role of symmetric tilt grain boundary (STGB) structures on the distribution of helium in nickel using molecular dynamics simulations. We investigate the family of STGBs specific to the 〈 110 〉 tilt axis. The present results indicate that accumulation of helium at the grain boundary may be modulated by details of grain boundary geometry. A plausible correlation between the grain boundary energy and misorientation with the accumulation and mobility of helium is proposed. Small clusters with up to 6 helium atoms show significant interstitial mobility in the nickel bulk, but also become sites for nucleation and grow of more stable helium clusters. High-energy GBs are found mainly populated with small helium clusters. The high mobility of small clusters along the GBs indicates the role of these GBs as fast two-dimensional channels for diffusion. In contrast, the accumulation of helium in large helium clusters at low-energy STGB creates a favorable environment for the formation of large helium bubbles, indicating a potential role for low-energy STGB in promoting helium-induced GB embrittlement.

Influence of TiC precipitation in austenitic stainless steel on strength, ductility and helium embrittlement

International Nuclear Information System (INIS)

Kesternich, W.; Matta, M.K.; Rothaut, J.

1984-01-01

Creep experiments were performed on 1.4970 (German DIN standard) and 316 (AISI standard) type austenitic steels after various thermomechanical pretreatments and after α-implantation. The microstructure introduced by the pretreatments was characterized by transmission electron microscopy and the behaviour of strength and ductility is correlated to the dislocation and precipitate distributions. He embrittlement can be suppressed in these simulation experiments when dispersive TiC precipitate distributions are produced by the proper pretreatments or are allowed to form during creep testing. It is shown that adequate pretreatment results in a significantly superior behaviour of the 1.4970 steel as compared to the 316 type steel in all three investigated properties, i.e. strength, ductility and resistance to He embrittlement. (orig.)

Helium effect on mechanical property of fusion reactor structural materials

International Nuclear Information System (INIS)

Yamamoto, Norikazu; Chuto, Toshinori; Murase, Yoshiharu; Nakagawa, Johsei

2004-01-01

High-energy neutrons produced in fusion reactor core caused helium in the structural materials of fusion reactors, such as blankets. We injected alpha particles accelerated by the cyclotron to the samples of martensite steel (9Cr3WVTaB). Equivalent helium doses injected to the sample is estimated to be up to 300 ppm, which were estimated to be equivalent to helium accumulation after the 1-year reactor operation. Creep tests of the samples were made to investigate helium embrittlement. There were no appreciable changes in the relation between the stresses and the rupture time, the minimum creep rate and the applied stress. Grain boundary effect by helium was not observed in ruptured surfaces. Fatigue tests were made for SUS304 samples, which contain helium up to 150 ppm. After 0.05 Hz cyclic stress tests, it was shown that the fatigue lifetime (cycles to rupture and extension to failure) are 1/5 in 150 ppm helium samples compared with no helium samples. The experimental results suggest martensite steel is promising for structural materials of fusion reactors. (Y. Tanaka)

Effects of exposure to high-temperature helium containing oxygen on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature

International Nuclear Information System (INIS)

Noda, T.; Okada, M.; Watanabe, R.

1980-01-01

The effects of exposure to helium containing oxygen of 0.1-115 vpm at 1000 0 C on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature were studied. The stress-relieved molybdenum specimen which was not recrystallized at test temperature showed the ductility after exposure to helium containing oxygen. The recrystallized molybdenum and TZM lost ductility after exposure to helium containing oxygen of 0.1-13 vpm in a few hours. The embrittlement of molybdenum was considered to be due to the grain boundary weakening. Molybdenum to which carbon was added seemed to hinder the grain boundary weakening by the oxygen contamination. Both stress-relieved and recrystallized TZM specimens picked up oxygen linearly with time of exposure to helium. The increase in oxygen content of TZM, which was considered to be caused by the internal oxidation of titanium and zirconium, results in the embrittlement of TZM. (orig.)

Microstructure and thermomechanical pretreatment effects on creep behaviour of helium-implanted DIN 1.4970 austenitic stainless steel

International Nuclear Information System (INIS)

Matta, M.K.; Kesternich, W.

1990-01-01

Microstructure investigations were carried out on unimplanted and 150 at ppm helium implanted foil specimens of DIN 1.4970 austenitic stainless steel after various thermomechanical pretreatments. Creep test were also carried out for both helium-implanted and unimplanted specimens at 700degC and 800degC. The strength, ductility and rupture time are correalted with the dislocation and precipitate distributions. Helium embrittlement can be reduced in these experiments when dispersive TiC precipitate distributions are produced by proper pretreatments or allowed to form during creep test. (author). 14 refs., 11 figs

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

International Nuclear Information System (INIS)

Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

2012-01-01

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

Energy Technology Data Exchange (ETDEWEB)

Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2012-07-15

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Radiation embrittlement of metals and alloys

International Nuclear Information System (INIS)

Wechsler, M.S.

1975-01-01

Three types of radiation embrittlement are identified: (1) radiation embrittlement in nominally ductile metals, (2) radiation embrittlement in metals that undergo a ductile-brittle transition, and (3) high-temperature grain boundary embrittlement. This paper deals with type (1) and, more briefly, type (2) radiation embrittlement. Radiation embrittlement in nominally ductile metals is characterized by the premature onset of plastic instability, which causes a sharp decrease in the macroscopic plastic strain that the material can sustain before necking (uniform strain) and breaking (fracture strain). Dislocation channeling seems to be largely responsible and experimental results are reviewed. The origin of dislocation channeling is discussed. Irradiated metals that exhibit a ductile-brittle transition show an increase in the transition temperature but the nature of the transition (shear to cleavage fracture) does not appear to be greatly altered. A key factor is the temperature dependence of yielding and how it is affected upon irradiation. Impurities exert an influence on the stability of radiation-produced defect clusters and thus can alter the amount of radiation embrittlement experienced upon irradiation at somewhat elevated temperatures. In general, radiation embrittlement appears to stem mostly from changes in plastic properties (particularly in the trend toward more dynamic and inhomogeneous plastic deformation) rather than from changes in the inherent fracture process. 63 references, 10 figures

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

Hardening Embrittlement and Non-Hardening Embrittlement of Welding-Heat-Affected Zones in a Cr-Mo Low Alloy Steel

Directory of Open Access Journals (Sweden)

Yu Zhao

2018-06-01

Full Text Available The embrittlement of heat affected zones (HAZs resulting from the welding of a P-doped 2.25Cr-1Mo steel was studied by the analysis of the fracture appearance transition temperatures (FATTs of the HAZs simulated under a heat input of 45 kJ/cm with different peak temperatures. The FATTs of the HAZs both with and without tempering increased with the rise of the peak temperature. However, the FATTs were apparently lower for the tempered HAZs. For the as-welded (untempered HAZs, the FATTs were mainly affected by residual stress, martensite/austenite (M/A islands, and bainite morphology. The observed embrittlement is a hardening embrittlement. On the other hand, the FATTs of the tempered HAZs were mainly affected by phosphorus grain boundary segregation, thereby causing a non-hardening embrittlement. The results demonstrate that the hardening embrittlement of the as-welded HAZs was more severe than the non-hardening embrittlement of the tempered HAZs. Consequently, a post-weld heat treatment should be carried out if possible so as to eliminate the hardening embrittlement.

Tensile properties of helium-injected V-15Cr-5Ti after irradiation in EBR-II

International Nuclear Information System (INIS)

Grossbeck, M.L.; Horak, J.A.

1985-01-01

Miniature specimens of V-15Cr-5Ti were prepared in the annealed condition and with 10, 20, and 30% cold work. The annealed specimens were cyclotron injected with helium and irradiated in sodium in EBR-II. The cold-worked specimens were irradiated in EBR-II but not helium injected. The specimens were irradiated at 400, 525, 625, and 700 0 C and received a fluence of 4.1 to 5.5 x 10 26 neutrons/m 2 (E > 0.1 meV). Tensile testing revealed very significant embrittlement as a result of the neutron irradiation but a much smaller change, mostly at 400 0 C, resulting from helium injection. 5 references, 9 figures, 2 tables

The helium effect at grain boundaries in Fe-Cr alloys: A first-principles study

Energy Technology Data Exchange (ETDEWEB)

Zemła, M.R., E-mail: marcin.zemla@wimpw.edu.pl [Faculty of Materials Science and Engineering, Warsaw University of Technology, Wołoska 141, 02-507 Warsaw (Poland); Wróbel, J.S.; Wejrzanowski, T. [Faculty of Materials Science and Engineering, Warsaw University of Technology, Wołoska 141, 02-507 Warsaw (Poland); Nguyen-Manh, D. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Kurzydłowski, K.J. [Faculty of Materials Science and Engineering, Warsaw University of Technology, Wołoska 141, 02-507 Warsaw (Poland)

2017-02-15

Helium is produced in the structural materials in nuclear power plants by nuclear transmutation following neutron irradiation. Since the solubility of helium in all metals is extremely low, helium tends to be trapped at defects such as vacancies, dislocations and grain boundaries, which cause material embrittlement. Density functional theory (DFT) calculations were performed in order to investigate the helium effect at grain boundaries (GBs) in iron-chromium alloys. Both cohesive energy and magnetic properties at symmetric Σ3(1 1 1) and Σ5(2 1 0) tilt Fe GBs are studied in the presence of Cr and He atoms. It is found that the presence of Cr atoms increases cohesive energy, at different He concentrations, and strongly influences magnetic properties at the GBs. The effect of the segregation energy of helium atom as a function of the different positions of Cr atoms located inside/outside a GB has been considered. Results of the present first-principles study enable one to clarify the role of Cr in understanding the helium effect in Fe-Cr-based alloys.

Simulating the CTR environment in the HVEM. Final report, September 1, 1975-August 31, 1984

International Nuclear Information System (INIS)

Jesser, W.A.

1984-01-01

The first high voltage electron microscope-ion accelerator facility in the world was established in 1976. Through use of this facility, helium embrittlement as revealed through intergranular fracture was shown to be independent of helium bubble content in the grain boundaries. This result is contrary to most current thinking and theories of helium embrittlement. The most likely mechanism of helium embrittlement is a plastic deformation assisted distribution of helium in finely dispersed helium-vacancy clusters. The key to overcoming helium embrittlement should lie in preventing the fine dispersion by providing benign sinks or treatments that agglomerate the helium

Power reactor embrittlement data base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1989-01-01

Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

Embrittlement data base, version 1

International Nuclear Information System (INIS)

Wang, J.A.

1997-08-01

The aging and degradation of light-water-reactor (LWR) pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel (RPV) materials depends on many different factors such as flux, fluence, fluence spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters and issues such as low-leakage-fuel management, possible life extension, and the need for annealing the pressure vessel. Large amounts of data from surveillance capsules and test reactor experiments, comprising many different materials and different irradiation conditions, are needed to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) is such a comprehensive collection of data resulting from merging version 2 of the Power Reactor Embrittlement Data Base (PR-EDB). Fracture toughness data were also integrated into Version 1 of the EDB. For power reactor data, the current EDB lists the 1,029 Charpy transition-temperature shift data points, which include 321 from plates, 125 from forgoings, 115 from correlation monitor materials, 246 from welds, and 222 from heat-affected-zone (HAZ) materials that were irradiated in 271 capsules from 101 commercial power reactors. For test reactor data, information is available for 1,308 different irradiated sets (352 from plates, 186 from forgoings, 303 from correlation monitor materials, 396 from welds and 71 from HAZs) and 268 different irradiated plus annealed data sets

Hydrogen embrittlement of steels: study and prevention

International Nuclear Information System (INIS)

Brass, A.M.; Chene, J.; Coudreuse, L.

2000-01-01

Hydrogen embrittlement of steels is one of the important reason of rupture of pieces in the industry (nuclear, of petroleum..). Indeed, there are a lot of situations which can lead to the phenomenon of hydrogen embrittlement: introduction of hydrogen in the material during the elaboration or during transformation or implementation processes (heat treatments, welding); use of steels when hydrogen or hydrogenated gaseous mixtures are present; hydrogen produced by electrolytic reactions (surface treatments, cathodic protection). The hydrogen embrittlement can appear in different forms which depend of a lot of parameters: material (state, composition, microstructure..); surrounding medium (gas, aqueous medium, temperature..); condition of mechanical solicitation (static, dynamic, cyclic..). The industrial phenomena which appear during cases of hydrogen embrittlement are more particularly described here. Several methods of steels studies are proposed as well as some possible ways for the prevention of hydrogen embrittlement risks. (O.M.)

Blistering and hydride embrittlement

International Nuclear Information System (INIS)

Louthan, M.R. Jr.

1975-01-01

The effects of hydrogen on the mechanical properties of metals have been categorized into several groups. Two of the groups, hydrogen blistering and hydride embrittlement, are reasonably well understood, and problems relating to their occurrence may be avoided if that understanding is used as a basis for selecting alloys for hydrogen service. Blistering and hydride embrittlement are described along with several techniques of materials selection and used to minimize their adverse effects. (U.S.)

Empirical Method to Estimate Hydrogen Embrittlement of Metals as a Function of Hydrogen Gas Pressure at Constant Temperature

Science.gov (United States)

Lee, Jonathan A.

2010-01-01

High pressure Hydrogen (H) gas has been known to have a deleterious effect on the mechanical properties of certain metals, particularly, the notched tensile strength, fracture toughness and ductility. The ratio of these properties in Hydrogen as compared to Helium or Air is called the Hydrogen Environment Embrittlement (HEE) Index, which is a useful method to classify the severity of H embrittlement and to aid in the material screening and selection for safety usage H gas environment. A comprehensive world-wide database compilation, in the past 50 years, has shown that the HEE index is mostly collected at two conveniently high H pressure points of 5 ksi and 10 ksi near room temperature. Since H embrittlement is directly related to pressure, the lack of HEE index at other pressure points has posed a technical problem for the designers to select appropriate materials at a specific H pressure for various applications in aerospace, alternate and renewable energy sectors for an emerging hydrogen economy. Based on the Power-Law mathematical relationship, an empirical method to accurately predict the HEE index, as a function of H pressure at constant temperature, is presented with a brief review on Sievert's law for gas-metal absorption.

Specificity in liquid metal induced embrittlement

CSIR Research Space (South Africa)

Fernandes, PJL

1996-12-01

Full Text Available One of the most intriguing features of liquid metal induced embrittlement (LMIE) is the observation that some liquid metal-solid metal couples are susceptible to embrittlement, while others appear to be immune. This is referred to as the specificity...

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.

Hydrogen embrittlement and galvanic corrosion of titanium alloys

Energy Technology Data Exchange (ETDEWEB)

Soh, Jeong Ryong; Jeong, Y. H.; Choi, B. K.; Baek, J. H.; Hwang, D. Y.; Choi, B. S.; Lee, D. J

2000-06-01

The material properties including the fracture behavior of titanium alloys used as a steam generator tube in SMART can be degraded de to the hydrogen embrittlement and the galvanic corrosion occurring as a result of other materials in contact with titanium alloys in a conducting corrosive environment. In this report the general concepts and trends of hydrogen embrittlement are qualitatively described to adequately understand and expect the fracture behavior from hydrogen within the bulk of materials and under hydrogen containing environments because hydrogen embrittlement may be very complicated process. And the characteristics of galvanic corrosion closely related to hydrogen embrittlement is qualitatively based on wimple electrochemical theory.

Hydrogen embrittlement and galvanic corrosion of titanium alloys

International Nuclear Information System (INIS)

Soh, Jeong Ryong; Jeong, Y. H.; Choi, B. K.; Baek, J. H.; Hwang, D. Y.; Choi, B. S.; Lee, D. J.

2000-06-01

The material properties including the fracture behavior of titanium alloys used as a steam generator tube in SMART can be degraded de to the hydrogen embrittlement and the galvanic corrosion occurring as a result of other materials in contact with titanium alloys in a conducting corrosive environment. In this report the general concepts and trends of hydrogen embrittlement are qualitatively described to adequately understand and expect the fracture behavior from hydrogen within the bulk of materials and under hydrogen containing environments because hydrogen embrittlement may be very complicated process. And the characteristics of galvanic corrosion closely related to hydrogen embrittlement is qualitatively based on wimple electrochemical theory

Effect of helium on creep and fatigue (MAT 11)

International Nuclear Information System (INIS)

Schroeder, H.

1991-03-01

This final report contains experimental results on mechanical properties (creep, fatigue, tensile) and microstructural investigations (SEM, TEM) of pre-implanted samples of steels or alloys. (AISI 316, AISI 316L, DIN 1.4970, JPCA 8206, DIN 1.4914; Incoloy 800H, Hastelloy X, DIN 1.4981, (Fe 0.49 Ni 0.51 ) 3 V, Fe17Ni17Cr, Fe15Ni15Cr, Nimonic PE 16, Ni8Si). Furthermore theoretical aspects and developed models and mechanisms for helium embrittlement are described. This report is presented in the form of an extended summary without figures. (MM)

Liquid and Solid Metal Embrittlement.

Science.gov (United States)

1981-09-05

example, embrittlement of AISI 4140 steel begins at T/T, - 0.75 for cadmium, and 0.85 for lead and tin environments (2). In a few cases, e.g. zinc...has recently proposed, however, that liquid zinc can penetrate to very near the tip of a sharp crack in 4140 steel, based upon both direct observation...long could be detected, was observed in delayed failure experi- ments on unnotched 4140 steel, in the quenched and tempered condi- tion, embrittled by

U.S. NRC Embrittlement Data Base (EDB)

International Nuclear Information System (INIS)

Pace, J.V.; Rosseel, T.M.; Wang, J.A.

1999-01-01

Large amounts of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) [I] is such a comprehensive collection of such data resulting from the merging of the Power Reactor Embrittlement Data Base (PR-EDB) [2] and the Test Reactor Embrittlement Data Base (TR-EDB) [3]. Fracture toughness data were also integrated into Version 1 of the EDB. The EDB data files are in dBASE format and can be accessed with a personal computer using the DOS or WINDOWS operating system. A utility program has been written to investigate radiation embrittlement using this data base. The utility program is used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to tit and plot Charpy impact data

Radiation embrittlement of PWR vessel supports

International Nuclear Information System (INIS)

Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

1989-01-01

Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

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

A reassessment of the effects of helium on Charpy impact properties of ferritic/martensitic steels

International Nuclear Information System (INIS)

Gelles, D.S.; Hamilton, M.L.; Hankin, G.L.

1998-01-01

To test the effect of helium on Charpy impact properties of ferritic/martensitic steels, two approaches are reviewed: quantification of results of tests performed on specimens irradiated in reactors with very different neutron spectra, and isotopic tailoring experiments. Data analysis can show that if the differences in reactor response are indeed due to helium effects, then irradiation in a fusion machine at 400 C to 100 dpa and 1000 appm He will result in a ductile to brittle transition temperature shift of over 500 C. However, the response as a function of dose and helium level is unlikely to be simply due to helium based on physical reasoning. Shear punch tests and microstructural examinations also support this conclusion based on irradiated samples of a series of alloys made by adding various isotopes of nickel in order to vary the production of helium during irradiation in HFIR. The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys. However, helium itself, up to 75 appm at over 7 dpa appears to have little effect on the mechanical properties of the alloys. This behavior is instead understood to result from complex precipitation response. The database for effects of helium on embrittlement based on nickel additions is therefore probably misleading and experiments should be redesigned to avoid nickel precipitation

Irradiation embrittlement mitigation

Energy Technology Data Exchange (ETDEWEB)

Torronen, K; Pelli, R; Planman, T; Valo, M [Technical Research Centre of Finland, Jyvaeskylae (Finland). Combustion and Thermal Engineering Lab.

1994-12-31

Mitigation methods for reducing the irradiation damage on pressure vessel materials are reviewed: load leakage loading schemes are commonly used in PWRs to mitigate reactor pressure vessel embrittlement; dummy assemblies have been applied in WWER 440-type and in some old western power plants, when exceptional fast embrittlement has been encountered; shielding of the pressure vessel has been developed, but is not in common use; pre-stressing the pressure vessel has been proposed for preventing PTS failures, but its applicability is not yet demonstrated. The large number of successful annealing treatments performed in WWER 440 type reactors as well as research on the effects of annealing treatments suggest applications for western PWRs. The emergency core cooling systems have been modified in WWER 440-type reactors in connection with other mitigation measures. (authors). 37 refs., 18 figs., 2 tabs.

Irradiation embrittlement mitigation

International Nuclear Information System (INIS)

Torronen, K.; Pelli, R.; Planman, T.; Valo, M.

1993-01-01

Mitigation methods for reducing the irradiation damage on pressure vessel materials are reviewed: load leakage loading schemes are commonly used in PWRs to mitigate reactor pressure vessel embrittlement; dummy assemblies have been applied in WWER 440-type and in some old western power plants, when exceptional fast embrittlement has been encountered; shielding of the pressure vessel has been developed, but is not in common use; pre-stressing the pressure vessel has been proposed for preventing PTS failures, but its applicability is not yet demonstrated. The large number of successful annealing treatments performed in WWER 440 type reactors as well as research on the effects of annealing treatments suggest applications for western PWRs. The emergency core cooling systems have been modified in WWER 440-type reactors in connection with other mitigation measures. (authors). 37 refs., 18 figs., 2 tabs

Effect of implanted helium on tensile properties and hardness of 9% Cr martensitic stainless steels

Science.gov (United States)

Jung, P.; Henry, J.; Chen, J.; Brachet, J.-C.

2003-05-01

Hundred micrometer thick specimens of 9% Cr martensitic steels EM10 and T91 were homogeneously implanted with He 4 to concentrations up to 0.5 at.% at temperatures from 150 to 550 °C. The specimens were tensile tested at room temperature and at the respective implantation temperatures. Subsequently the fracture surfaces were analysed by scanning electron microscopy and some of the specimens were examined in an instrumented hardness tester. The implanted helium caused hardening and embrittlement which both increased with increasing helium content and with decreasing implantation temperature. Fracture surfaces showed intergranular brittle appearance with virtually no necking at the highest implantation doses, when implanted below 250 °C. The present tensile results can be scaled to tensile data after irradiation in spallation sources on the basis of helium content but not on displacement damage. An interpretation of this finding by microstructural examination is given in a companion paper [J. Nucl. Mater., these Proceedings].

Radiation embrittlement of WWER-1000 reactor vessel steels

International Nuclear Information System (INIS)

Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

2001-01-01

Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

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

Crystal orientation effects on helium ion depth distributions and adatom formation processes in plasma-facing tungsten

International Nuclear Information System (INIS)

Hammond, Karl D.; Wirth, Brian D.

2014-01-01

We present atomistic simulations that show the effect of surface orientation on helium depth distributions and surface feature formation as a result of low-energy helium plasma exposure. We find a pronounced effect of surface orientation on the initial depth of implanted helium ions, as well as a difference in reflection and helium retention across different surface orientations. Our results indicate that single helium interstitials are sufficient to induce the formation of adatom/substitutional helium pairs under certain highly corrugated tungsten surfaces, such as (1 1 1)-orientations, leading to the formation of a relatively concentrated layer of immobile helium immediately below the surface. The energies involved for helium-induced adatom formation on (1 1 1) and (2 1 1) surfaces are exoergic for even a single adatom very close to the surface, while (0 0 1) and (0 1 1) surfaces require two or even three helium atoms in a cluster before a substitutional helium cluster and adatom will form with reasonable probability. This phenomenon results in much higher initial helium retention during helium plasma exposure to (1 1 1) and (2 1 1) tungsten surfaces than is observed for (0 0 1) or (0 1 1) surfaces and is much higher than can be attributed to differences in the initial depth distributions alone. The layer thus formed may serve as nucleation sites for further bubble formation and growth or as a source of material embrittlement or fatigue, which may have implications for the formation of tungsten “fuzz” in plasma-facing divertors for magnetic-confinement nuclear fusion reactors and/or the lifetime of such divertors.

Updated embrittlement trend curve for reactor pressure vessel steels

International Nuclear Information System (INIS)

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

2003-01-01

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

The Test Reactor Embrittlement Data Base (TR-EDB)

International Nuclear Information System (INIS)

Stallmann, F.W.; Kam, F.B.K.; Wang, J.A.

1993-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to assure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB) whose second version is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in fall of 1993 and contains published results from laboratories in many countries. Data collection will continue and further updates will be published

Compatibility of vanadium alloys with reactor-grade helium for fusion reactor applications

International Nuclear Information System (INIS)

Bell, G.E.C.; Bishop, P.S.

1993-01-01

Tests were conducted to determine the compatibility of vanadium alloys with reactor-grade helium and to define the helium gas chemistry requirements for fusion reactors, miniature tensile specimens of V-5Cr-5Ti. V-10Cr-5Ti, and V-12.5Cr-5 Ti were exposed in a once-through system to helium with 70 vppm-H 2 (measured oxygen partial pressures of 10 -12 atm) and bottle helium (measured oxygen partial pressures of -4 atm) between 500 and 700 degree C for up to 1008 h. The weight changes in the specimens were recorded. The helium-exposed specimens were tensile tested, and the effects of exposure on mechanical properties were assessed. Exposure between 500 and 700 degree C for 1008 h in He+70 vppm-H 2 resulted in complete embrittlement of all the alloys in room temperature tensile tests. The fracture mode was primarily cleavage, probably caused by a hydrogen-induced shift in the ductile to brittle transition temperature (DBTT). Weight gains increased with temperature and were largest for the V-5Cr-5Ti alloy. Specimens exposed for 531 h between 500 and 700 degree C in bottle He exhibited two distinct fracture morphologies on the fracture surfaces. Brittle cleavage around the edges of specimens gave way to ductile dimpling in the center of the specimens. The brittle region around the periphery of the specimen is most likely the highest vanadium oxide. V 2 O 5

Modeling of helium effects in metals: High temperature embrittlement

International Nuclear Information System (INIS)

Trinkaus, H.

1985-01-01

The effects of helium on swelling, creep rupture and fatigue properties of fusion reactor materials subjected to (n,α)-reactions and/or direct α-injection, are controlled by bubble formation. The understanding of such effects requires therefore the modeling of (1) diffusional reactions of He atoms with other defects; (2) nucleation and growth of He bubbles; (3) transformation of such bubbles into cavities under continuous He generation and irradiation or creep stress. The present paper is focussed on the modeling of the (coupled) high temperature bubble nucleation and growth processes within and on grain boundaries. Two limiting cases are considered: di-atomic nucleation described by the simplest possible sets of rate equations, and multi-atomic nucleation described by classical nucleation theory. Scaling laws are derived which characterize the dependence of the bubble densities upon time (He-dose), He generation rate and temperature. Comparison with experimental data of AISI 316 SS α-implanted at temperatures around 1000 K indicates bubble nucleation of the multi-atomic type. The nucleation and growth models are applied to creep tests performed during α-implantation suggesting that in these cases gas driven bubble growth is the life time controlling mechanism. The narrow (creep stress/He generation rate) range of this mechanism in a mechanism map constructed from these tests indicates that in many reactor situations the time to rupture is probably controlled by stress driven cavity growth rather than by gas driven bubble growth. (orig.)

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)

The liquid metal embrittlement of iron and ferritic steels in sodium

International Nuclear Information System (INIS)

Hilditch, J.P.; Hurley, J.R.; Tice, D.R.; Skeldon, P.

1995-01-01

The liquid metal embrittlement of iron and A508 III, 21/4Cr-1Mo and 15Mo3 steels in sodium at 200-400 o C has been studied, using dynamic straining at 10 -6 s -1 , in order to investigate the roles of microstructure and composition. The steels comprised bainitic, martensitic, tempered martensitic and ferritic/pearlitic microstructures. All materials were embrittled by sodium, the embrittlement being associated generally with quasicleavage on fracture surfaces. Intergranular cracking was also found with martensitic and ferritic/pearlitic microstructures. The susceptibility to embrittlement was greater in higher strength materials and at higher temperatures. The embrittlement was similar to that encountered previously in 9Cr steel, which depends upon the presence of non-metallic impurities in the sodium. (author)

Embrittlement and life prediction of aged duplex stainless steel

International Nuclear Information System (INIS)

Kuwano, Hisashi

1996-01-01

The stainless steel, for which the durability for long term in high temperature corrosive environment is demanded, is a complex plural alloy. Cr heightens the oxidation resistance, Ni improves the ductility and impact characteristics, Si improves the fluidity of the melted alloy and heightens the resistance to stress corrosion cracking, and Mo suppresses the pitting due to chlorine ions. These alloy elements are in the state of nonequilibrium solid solution in Fe base at practical temperature, and cause aging phenomena such as segregation, concentration abnormality and precipitation during the use for long term. The characteristics of stainless steel deteriorate due to this. Two-phase stainless cast steel, the example of the embrittlement of the material for an actual machine, the accelerated test of embrittlement, the activation energy for embrittlement, and as the mechanism of aging embrittlement, the spinodal decomposition of ferrite, the precipitation of G phase and the precipitation of carbides and nitrides are described. Also in the welded parts of austenitic stainless steel, delta-ferrite is formed during cooling, therefore, the condition is nearly same as two-phase stainless steel, and the embrittlement due to long term aging occurs. (K.I.)

Ab initio Investigation of Helium in Vanadium Oxide Nanoclusters

Science.gov (United States)

Danielson, Thomas; Tea, Eric; Hin, Celine

Nanostructured ferritic alloys (NFAs) are strong candidate materials for the next generation of fission reactors and future fusion reactors. They are characterized by a large number density of oxide nanoclusters dispersed throughout a BCC iron matrix, where current oxide nanoclusters are primarily comprised of Y-Ti-O compounds. The oxide nanoclusters provide the alloy with high resistance to neutron irradiation, high yield strength and high creep strength at the elevated temperatures of a reactor environment. In addition, the oxide nanoclusters serve as trapping sites for transmutation product helium providing substantially increased resistance to catastrophic cracking and embrittlement. Although the mechanical properties and radiation resistance of the existing NFAs is promising, the problem of forming large scale reactor components continues to present a formidable challenge due to the high hardness and unpredictable fracture behavior of the alloys. An alternative alloy has been previously proposed and fabricated where vanadium is added in order to form vanadium oxide nanoclusters that serve as deflection sites for crack propagation. Although experiments have shown evidence that the fracture behavior of the alloys is improved, it is unknown whether or not the vanadium oxide nanoclusters are effective trapping sites for helium. We present results obtained using density functional theory investigating the thermodynamic stability of helium with the vanadium oxide matrix to make a comparison of trapping effectiveness to traditional Y-Ti-O compounds.

Power Reactor Embrittlement Data Base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1990-01-01

Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: (1) to compile and to verify the quality of the PR-EDB; (2) to provide user-friendly software to access and process the data; (3) to explore or confirm embrittlement prediction models; and (4) to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. To achieve these goals, the data base architecture was designed after much discussion and planning with prospective users, namely, material scientists and members of the research staff. The current compilation of the PR-EDB (Version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points for 110 different irradiated base materials and 161 data points for 79 different welds. Results from heat-affected zone materials are also listed. The time and effort required to process and evaluate different types of data in the PR-EDB have been drastically reduced from previous data bases. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of PR-EDB and will be supplementing the data base with additional data and documentation

Mechanisms of liquid-metal embrittlement

International Nuclear Information System (INIS)

Popovich, V.V.

1979-01-01

The mechanism of the embrittlement of metals and alloys during deformation in contact with liquid metals are discussed. With 20Kh13 steel in a Pb-Sn melt and polycrystalline Al in the presence of various mercury solutions a.s examples, considered are the three main processes - adsorption, corrosion (dissolution), formation of new phases which cause the disintegration of materials under the action of liquid-metallic media. Presented are data on plastic ductile and strength properties of the above materials in the presence of liquid-metallic media. A model is described that takes into account the effect of the medium upon the plastic deformation and the part the medium plays in liquid-metallic embrittlement

Proposal of guideline for bonding to prevention of hydrogen embrittlement at Ta/Zr bond interface. Hydrogen embrittlement in SUS304ULC/Ta/Zr explosive bonded joint

International Nuclear Information System (INIS)

Saida, Kazuyoshi; Fujimoto, Tetsuya; Nishimoto, Kazutoshi

2010-01-01

The occurrence condition of hydrogen embrittlement cracking at Ta/Zr bond interface was investigated with respect to the hydrogen content and applied stress in order to propose a guideline for the explosive bonding procedure to prevention of hydrogen embrittlement. Hydrogen charging test was conducted for SUS304ULC/Ta/Zr explosive bonded joints applied the different flexural strains. A hydrogen embrittlement crack occurred in the Zr substrate at Ta/Zr bond interface after hydrogen charging, and it was initiated at shorter charging times when the augmented strain was increased. The occurrence condition of hydrogen embrittlement cracking at Ta/Zr bond interface was shifted to lower stress and hydrogen content with an increase in the amount of explosive during bonding. It was suggested that hydrogen embrittlement in Ta/Zr explosive bonded joint could be inhibited by reducing the initial hydrogen content in Ta substrate less than approx. 5 ppm. (author)

Cladding embrittlement during postulated loss-of-coolant accidents.

Energy Technology Data Exchange (ETDEWEB)

Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

2008-07-31

The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

Estimation of embrittlement damage risk at neutron embrittled vessel constructions

International Nuclear Information System (INIS)

Staevski, K.; Madzharov, D.; Detistov, P.; Petrova, T.

1998-01-01

In this work a methodology based on Damage mechanics criteria is proposed. This methodology serves for probability assessment of the brittle damage risk for the neutron embrittled vessel elements. The developed methodology is realised in RISK code and has been verified on the base of tough reliability of the pressure vessel, 'Kozloduy' NPP Unit 2. This investigation has been carried out at the given parameters of the possible defects on the vessel's weld 4 taking into account requirements of the western and Russian standards. The obtained values for ductile to brittle transition temperatures, defining the equipment life-time in the presence of maximal defect, are in good consistence with the experimentally determined ones. The analyses of results show that the pressure vessel of 'Kozloduy' NPP Unit 2 has got a high level of reliability from brittle damage risk point of view and that the western standards give more conservative evaluation. On the bases of the results a conclusion is made that the developed methodology enables analysing the influence of possible defects in the neutron embrittled elements on their to reliability and their remained life-time

Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

International Nuclear Information System (INIS)

Reymann, G.A.

1979-02-01

A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

Simulation of He embrittlement at grain boundaries in bcc transition metals

Energy Technology Data Exchange (ETDEWEB)

Suzudo, Tomoaki, E-mail: suzudo.tomoaki@jaea.go.jp; Yamaguchi, Masatake

2015-10-15

To investigate what atomic properties largely determine vulnerability to He embrittlement at grain boundaries (GB) of bcc metals, we introduce a computational model composed of first principles density functional theory and a He segregation rate theory model. Predictive calculations of He embrittlement at the first wall of the future DEMO fusion concept reactor indicate that variation in the He embrittlement originated not only from He production rate related to neutron irradiation, but also from the He segregation energy at the GB that has a systematic trend in the periodic table. - Highlights: • We modeled He grain boundary (GB) segregation of bcc transition metals using first-principles-based rate theory. • We established the quantitative relation between He embrittlement and He segregation using GB cohesive energy. • He embrittlement was strongly dependent on He segregation energy at the GB that has a systematic trend in the periodic table.

Simulation of He embrittlement at grain boundaries in bcc transition metals

International Nuclear Information System (INIS)

Suzudo, Tomoaki; Yamaguchi, Masatake

2015-01-01

To investigate what atomic properties largely determine vulnerability to He embrittlement at grain boundaries (GB) of bcc metals, we introduce a computational model composed of first principles density functional theory and a He segregation rate theory model. Predictive calculations of He embrittlement at the first wall of the future DEMO fusion concept reactor indicate that variation in the He embrittlement originated not only from He production rate related to neutron irradiation, but also from the He segregation energy at the GB that has a systematic trend in the periodic table. - Highlights: • We modeled He grain boundary (GB) segregation of bcc transition metals using first-principles-based rate theory. • We established the quantitative relation between He embrittlement and He segregation using GB cohesive energy. • He embrittlement was strongly dependent on He segregation energy at the GB that has a systematic trend in the periodic table.

Reactor pressure vessel embrittlement: Insights from neural network modelling

Science.gov (United States)

Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

2018-04-01

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

Tritium Decay Helium-3 Effects in Tungsten

Energy Technology Data Exchange (ETDEWEB)

Shimada, M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Merrill, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

A critical challenge for long-term operation of ITER and beyond to a Demonstration reactor (DEMO) and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to steady-state/transient heat fluxes and intense neutral/ion particle fluxes under the extreme fusion nuclear environment, while at the same time minimizing in-vessel tritium inventories and permeation fluxes into the PFC’s coolant. Tritium will diffuse in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [1,2]. Tritium decay into helium-3 may also play a major role in microstructural evolution (e.g. helium embrittlement) in tungsten due to relatively low helium-4 production (e.g. He/dpa ratio of 0.4-0.7 appm [3]) in tungsten. Tritium-decay helium-3 effect on tungsten is hardly understood, and its database is very limited. Two tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) were exposed to high flux (ion flux of 1.0x1022 m-2s-1 and ion fluence of 1.0x1026 m-2) 0.5%T2/D2 plasma at two different temperatures (200, and 500°C) in Tritium Plasma Experiment (TPE) at Idaho National Laboratory. Tritium implanted samples were stored at ambient temperature in air for more than 3 years to investigate tritium decay helium-3 effect in tungsten. The tritium distributions on plasma-exposed was monitored by a tritium imaging plate technique during storage period [4]. Thermal desorption spectroscopy was performed with a ramp rate of 10°C/min up to 900°C to outgas residual deuterium and tritium but keep helium-3 in tungsten. These helium-3 implanted samples were exposed to deuterium plasma in TPE to investigate helium-3 effect on deuterium behavior in tungsten. The results show that tritium surface concentration in 200°C sample decreased to 30 %, but tritium surface concentration in 500°C sample did not alter over the 3 years storage period, indicating possible tritium

Intrinsic ductility and environmental embrittlement of binary Ni3Al

International Nuclear Information System (INIS)

George, E.P.; Liu, C.T.; Pope, D.P.

1993-01-01

Polycrystalline, B-free Ni 3 Al (23.4 at.% Al), produced by cold working and recrystallizing a single crystal, exhibits room temperature tensile ductilities of 3-5% in air and 13-16% in oxygen. These ductilities are considerably higher than anything previously reported, and demonstrate that the 'intrinsic' ductility of Ni 3 Al is much higher than previously thought. They also show that the moisture present in ordinary ambient air can severely embrittle Ni 3 Al (ductility decreasing from a high of 16% in oxygen to a low of 3% in air). Fracture is predominantly intergranular in both air and oxygen. This indicates that, while moisture can further embrittle the GBs in Ni 3 Al, they persist as weak links even in the absence of environmental embrittlement. However, they are not 'intrinsically brittle' as once thought, since they can withstand relatively large plastic deformations prior to fracture. Because B essentially eliminates environmental embrittlement in Ni 3 Al - and environmental embrittlement is a major cause of poor ductility in B-free Ni 3 Al - it is concluded that a significant portion of the so-called B effect must be related to suppression of moisture-induced environmental embrittlement. However, since B-doped Ni 3 Al fractures transgranularly, whereas B-free Ni 3 Al fractures predominantly intergranularly, B must have the added effect that it strengthens the GBs. A comparison with the earlier work on Zr-doped Ni 3 Al shows that Zr improves the ductility of Ni 3 Al, both in air and (and even more dramatically) in oxygen. While the exact mechanism of this ductility improvement is not clear at present, Zr appears to have more of an effect on (enhancing) GB strength than on (suppressing) environmental embrittlement

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-24

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

Low temperature hydrogen embrittlement of niobium. II. Microscopic observations

International Nuclear Information System (INIS)

Grossbeck, M.L.; Birnbaum, H.K.

1977-01-01

The detailed, microscopic processes which occur during the hydrogen embrittlement of pure Nb are examined using in situ SEM crack propagation studies, SEM fractography, electron diffraction and ion probe methods. These results show that the fracture process occurs in a stress induced NbH hydride phase which forms in front of the propagating crack. The experimental results are in good agreement with the stress induced hydride embrittlement mechanism which is discussed. The thermodynamics of precipitation of hydrides under external stress is discussed and calculations are presented for the stress effects on the α-β solvus temperatures. These are related to the embrittlement process and evidence is presented to support the calculated stress effects on the solvus temperature

Study of helium diffusion, implanted at a cyclotron, in face-centered cubic metals: Au, Ag and Al

International Nuclear Information System (INIS)

Sciani, V.

1985-01-01

Helium in metals is produced by nuclear reactions of energetic particles. In nuclear technology the interest on helium in metals is import, due to its production by (n, α) reaction. Because helium has extremely low solubility in metals, the precipitation in the form of filled bubbles at elevated temperatures occurs, which have detrimental effects on mechanical properties and may limit the lifetime of structural components. One typical example is the high temperature embrittlement. The nucleation and growth of the bubbles strongly depends on the mobility of the helium. This work presents the study of helium diffusion in Au, Ag and Al at temperatures above room temperature. The helium created by (n, α) reactions has been simulated by homogeneous alpha particles implantation in cyclotron, at room temperature, in specimens of thicknesses between 5 and 50 μm and helium concentration between 10 -3 to 10 ppm. After implantation, the specimens were dropped in a furnace in a UHV-chamber and the diffusion was measured by observing the He-release during linear and isothermal annealings. The occurence of free diffusion was comparing the dependence of release kinetics on helium concentration, sample thickness, time and heating rate to diffusion theory and is clearly separeted from agglomeration process. The diffusion constants of helium in Au, Ag and Al follow an Arrhenius behavior, with: Au:D o =10 -1.0 cm 2 /s ΔH=1.70eV Ag:D 0 =10 -1.2 cm 2 /s ΔH=1.51eV Al:D o =10 +0.5 cm 2 /s ΔH=1.40eV. The results are compared to self-diffusion and to the diffusion of other gases in these metals. Comparison with theoretical estimates favours the vacancy mechanism for helium diffusion in Au, Ag and Al. (author) [pt

Neutron-induced hydrogen and helium production in iron

Energy Technology Data Exchange (ETDEWEB)

Haight, Robert C.

2004-01-01

In support of the Advanced Fuel Cycle Initiative, cross sections for hydrogen and helium production by neutrons are being investigated on structural materials from threshold to 100 MeV with the continuous-in-energy spallation neutron source at the Los Alamos Neutron Science Center (LANSCE). The present measurements are for elemental iron. The results are compared with values from the ENDF/B-VI library and its extension with LA150 evaluations. For designs in the Advanced Fuel Cycle Initiative, structural materials will be subjected to very large fluences of neutrons, and the selection of these materials will be guided by their resistance to radiation damage. The macroscopic effects of radiation damage result both from displacement of atoms in the materials as well as nuclear transmutation. We are studying the production of hydrogen and helium by neutrons, because these gases can lead to significant changes in materials properties such as embrittlement and swelling. Our experiments span the full range from threshold to 100 MeV. The lower neutron energies are those characteristic of fission neutrons, whereas the higher energies are relevant for accelerator-based irradiation test facilities. Results for the nickel isotopes, {sup 58,60}Ni, have been reported previously. The present studies are on natural iron.

Re-examining reactor vessel embrittlement at Chooz A

International Nuclear Information System (INIS)

Guilleret, J.-C.

1988-01-01

The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

Power Reactor Embrittlement Data Base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1990-01-01

Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

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.)

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)

Grain boundary embrittlement and cohesion enhancement in copper

Energy Technology Data Exchange (ETDEWEB)

Paxton, Anthony; Lozovoi, Alexander [Atomistic Simulation Centre, Queen' s University Belfast, BT7 1NN (United Kingdom); Schweinfest, Rainer [Science+Computing ag, Hagellocher Weg 71-5, 720270 T ubingen (Germany); Finnis, Michael [Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

2008-07-01

There has been a long standing debate surrounding the mechanism of grain boundary embrittlement and cohesion enhancement in metals. Embrittlement can lead to catastrophic failure such as happened in the Hinkley Point disaster, or indeed in the case of the Titanic. This kind of embrittlement is caused by segregation of low solubility impurities to grain boundaries. While the accepted wisdom is that this is a phenomenon driven by electronic or chemical factors, using language such as charge transfer and electronegativity difference; we believe that in copper, at least, both cohesion enhancement and reduction are caused by a simple size effect. We have developed a theory that allows us to separate unambiguously, if not uniquely, chemical and structural factors. We have studied a large number of solutes in copper using first principles atomistic simulation to support this argument, and the results of these calculations are presented here.

Different approaches to estimation of reactor pressure vessel material embrittlement

Directory of Open Access Journals (Sweden)

V. M. Revka

2013-03-01

Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

Hydrogen embrittlement susceptibility of laser-hardened 4140 steel

Energy Technology Data Exchange (ETDEWEB)

Tsay, L.W.; Lin, Z.W. [Nat. Taiwan Ocean Univ., Keelung (Taiwan). Inst. of Mater. Eng.; Shiue, R.K. [Institute of Materials Sciences and Engineering, National Dong Hwa University, Hualien, Taiwan (Taiwan); Chen, C. [Institute of Materials Sciences and Engineering, National Taiwan University, Taipei, Taiwan (Taiwan)

2000-10-15

Slow strain rate tensile (SSRT) tests were performed to investigate the susceptibility to hydrogen embrittlement of laser-hardened AISI 4140 specimens in air, gaseous hydrogen and saturated H{sub 2}S solution. Experimental results indicated that round bar specimens with two parallel hardened bands on opposite sides along the loading axis (i.e. the PH specimens), exhibited a huge reduction in tensile ductility for all test environments. While circular-hardened (CH) specimens with 1 mm hardened depth and 6 mm wide within the gauge length were resistant to gaseous hydrogen embrittlement. However, fully hardened CH specimens became susceptible to hydrogen embrittlement for testing in air at a lower strain rate. The strength of CH specimens increased with decreasing the depth of hardened zones in a saturated H{sub 2}S solution. The premature failure of hardened zones in a susceptible environment caused the formation of brittle intergranular fracture and the decrease in tensile ductility. (orig.)

Influence of a cyclic load on the embrittlement kinetics of alloys by the example of the 475 C embrittlement of duplex steel and the dynamic embrittlement of a nickel base alloy; Einfluss einer zyklischen Belastung auf die Versproedungskinetik von Legierungen am Beispiel der 475 C-Versproedung von Duplexstahl und der dynamischen Versproedung einer Nickelbasislegierung

Energy Technology Data Exchange (ETDEWEB)

Wackermann, Ken

2015-07-07

The objective of this study was to investigate the dependence of high temperature embrittlement mechanisms on high temperature fatigue and vice versa. As model embrittlement mechanisms the 475 C Embrittlement of ferritic austenitic duplex stainless steel (1.4462) and the Dynamic Embrittlement of nickel-based superalloys (IN718) were selected. The 475 C Embrittlement is a thermally activated decomposition of the ferritic phase which hardens the material. In contrast to this a cyclic plastic deformation weakens the steel by a deformation-induced dissolution of the decomposition. Fatigue tests with different frequencies, loading amplitudes at room temperature and at 475 C with Duplex Stainless Steel in different states of embrittlement show that the ongoing 475 C Embrittlement and the deformation-induced dissolution are competing mechanisms. It depends on the frequency, the loading amplitude and the temperature which mechanism is dominant. Applying the model of the yield stress distribution function to the hysteresis branches of the fatigue tests allows an analysis of the fatigue behaviour of each phase individually. This analysis shows that the global fatigue behaviour for the test conditions applied in this study is mainly controlled by the ferritic phase. According to the existing understanding of Dynamic Embrittlement it is an oxygen grain boundary diffusion arising by tensile stress at elevated temperatures with the result of a fast intercrystalline crack propagation. In reference tests under vacuum conditions without oxygen grain boundary diffusion, a slow transcrystalline fracture appears. To analyse the Dynamic Embrittlement, the crack propagation was tested at 650 C with different frequencies and superimposed hold times in the fatigue cycle at maximum stress. The results shows that the existing model of Dynamic Embrittlement needs to be adapted to the effects of cyclic plastic deformation. In hold times, the oxygen grain boundary diffusion in front of the

The effect of helium, radiation damage and irradiation temperature on the mechanical properties of beryllium

Energy Technology Data Exchange (ETDEWEB)

Fabritsiev, S.A. [D.V. Efremov Scientific Research Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S.

1998-01-01

In this work different RF beryllium grades were irradiated in the BOR-60 reactor to a dose of {approx}5-10 dpa at irradiation temperatures 350, 420, 500, 800degC. Irradiation at temperatures of 350-400degC is shown to result in Be hardening due to the accumulation of radiation defect complexes. Hardening is accompanied with a sharp drop in plasticity at T{sub test} {<=} 300degC. A strong anisotropy in plasticity has been found at a mechanical testing temperature of 400degC and this parameter may be preferable when the samples are cut crosswise to the pressing direction. High-temperature irradiation (T{sub irr} = 780degC) gives rise to large helium pores over the grain boundaries and smaller pores in the grain body. Fracture is brittle and intercrystallite at T{sub test} {>=} 600degC. Helium embrittlement is accompanied as well with a drop in the Be strength properties. (author)

Study of helium behaviour in body-centered cubic structures for new nuclear reactor generations: experimental approach in well characterized materials

International Nuclear Information System (INIS)

Gorondy-Novak, Sofia Maria

2017-01-01

The presence of helium produced during the operation of future fast reactors and fusion reactors in core structural materials induces a deterioration of their mechanical properties (hardening, swelling, embrittlement). In order to pursue the development of the metallic structural alloys, it is necessary to comprehend the He interaction with the metal lattice thus the point in common is the study of the metallic components with body-centered cubic structure (bcc) of future alloys, such as iron and/or vanadium. Ion implantation of ions "4He was employed with the aim of simulating the damaging effects associated with the helium accumulation, the point defects' creation (vacancies, self-interstitials) and the He cluster formation in future reactors. Helium evolution in pure iron and pure vanadium has been revealed from the point of view of the trapping sites' nature and well as the helium migration mechanisms and the nucleation/growth of bubbles. These phenomena were studied by coupling different complementary techniques. Despite of the fact that some mechanisms involved seem to be similar for both bcc metals, the comparison between the helium behavior in iron and vanadium shows certain differences. Microstructural defects, including grain boundaries and implanted helium concentration (dose) in both bcc metals will play significant roles on the helium behavior at high temperature. The acquired experimental data coupled with simulation methods contribute to the future development in terms of kinetic and thermodynamic data management of helium behavior in the metal components of the alloys of nuclear interest. (author) [fr

Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

International Nuclear Information System (INIS)

Blom, F.J.

2007-01-01

Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

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

International Nuclear Information System (INIS)

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

1995-05-01

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

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)

The low-temperature aging embrittlement in a 2205 duplex stainless steel

International Nuclear Information System (INIS)

Weng, K.L.; Chen, H.R.; Yang, J.R.

2004-01-01

The effect of isothermal treatment (at temperatures ranging between 400 and 500 deg. C) on the embrittlement of a 2205 duplex stainless steel (with 45 ferrite-55 austenite, vol.%) has been investigated. The impact toughness and hardness of the aged specimens were measured, while the corresponding fractography was studied. The results show that the steel is susceptible to severe embrittlement when exposed at 475 deg. C; this aging embrittlement is analogous with that of the ferritic stainless steels, which is ascribed to the degenerated ferrite phase. High-resolution transmission electron microscopy reveals that an isotropic spinodal decomposition occurred during aging at 475 deg. C in the steel studied; the original δ-ferrite decomposed into a nanometer-scaled modulated structure with a complex interconnected network, which contained an iron-rich BCC phase (α) and a chromium-enriched BCC phase (α'). It is suggested that the locking of dislocations in the modulated structure leads to the severe embrittlement

Hydrogen embrittlement due to hydrogen-inclusion interactions

International Nuclear Information System (INIS)

Yu, H.Y.; Li, J.C.M.

1976-01-01

Plastic flow around inclusions creates elastic misfit which attracts hydrogen towards the regions of positive dilatation. Upon decohesion of the inclusion-matrix interface, the excess hydrogen escapes into the void and can produce sufficient pressure to cause void growth by plastic deformation. This mechanism of hydrogen embrittlement can be used to understand the increase of ductility with temperature, the decrease of ductility with hydrogen content, and the increase of ductility with the ultimate strength of the matrix. An examination of the effect of the shape of spheroid inclusion reveals that rods are more susceptible to hydrogen embrittlement than disks. The size of the inclusion is unimportant while the volume fraction of inclusions plays the usual role

Development of neutron irradiation embrittlement correlation of reactor pressure vessel materials of light water reactors

International Nuclear Information System (INIS)

Soneda, Naoki; Dohi, Kenji; Nomoto, Akiyoshi; Nishida, Kenji; Ishino, Shiori

2007-01-01

A large amount of surveillance data of the RPV embrittlement of the Japanese light water reactors have been compiled since the current Japanese embrittlement correlation has been issued in 1991. Understanding on the mechanisms of the embrittlement has also been greatly improved based on both experimental and theoretical studies. CRIEPI and the Japanese electric power utilities have started research project to develop a new embrittlement correlation method, where extensive study of the microstructural analyses of the surveillance specimens irradiated in the Japanese commercial reactors has been conducted. The new findings obtained from the experimental study are that the formation of solute-atom clusters with little or no copper is responsible for the embrittlement in low-copper materials, and that the flux effect exists especially in high-copper materials and this is supported by the difference in the microstructure of the high-copper materials irradiated at different fluxes. Based on these new findings, a new embrittlement correlation method is formulated using rate equations. The new methods has higher prediction capability than the current Japanese embrittlement correlation in terms of smaller standard deviation as well as smaller mean value of the prediction error. (author)

Studies on the effects of helium on the microstructural evolution of V-3.8Cr-3.9Ti

Energy Technology Data Exchange (ETDEWEB)

Doraiswamy, N.; Kestel, B.; Alexander, D.E. [Argonne National Labs., IL (United States)

1997-04-01

The favorable physical and mechanical properties of V-3.8Cr-3.9Ti (wt.%), when subjected to neutron irradiation, has lead to considerable attention being focused on it for use in fusion reactor structural applications. However, there is limited data on the effects of helium on physical and mechanical properties of this alloy. Understanding these effects are important since helium will be generated by direct {alpha}-injection or transmutation reactions in the fusion environment, typically at a rate of {approx}5 appm He/dpa. Helium has been shown to cause substantial embrittlement, even at room temperature in vanadium and its alloys. Recent simulations of the fusion environment using the Dynamic Helium Charging Experiments (DHCE) have also indicated that the mechanical properties of vanadium alloys are altered by the presence of helium in post irradiation tests performed at room temperature. While the strengths were lower, room temperature ductilities of the DHCE specimens were higher than those of non-DHCE specimens. These changes have been attributed to the formation of different types of hardening centers in these alloys due to He trapping. Independent thermal desorption experiments suggest that these hardening centers may be associated with helium-vacancy-X (where X = O, N, and C) complexes. These complexes are stable below 290{degrees}C and persist at room temperature. However, there has been no direct microstructural evidence correlating the complexes with irradiation effects. An examination of the irradiation induced microstructure in samples preimplanted with He to different levels would enable such a correlation.

Lifetime embrittlement of reactor core materials

International Nuclear Information System (INIS)

Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G.; White, C.J.

1994-08-01

Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10 24 n/M 2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K IC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K IC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

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

Survey of irradiation embrittlement effects on the mechanical properties of alloyed steels

International Nuclear Information System (INIS)

Gillemot, F.

1992-01-01

In the everyday engineering practice the neutron irradiation embrittlement of the PWR wall materials is measured by empirical methods like Charpy impact testing. New developments in fracture mechanics are given better material characteristics. The use of Absorbed Specific Fracture Energy Measured on tensile bars is a promising way to solve the problem. On the other hand the IAEA runs coordinated research program to correlate the chemical analysis with the rate of the neutron embrittlement. Better understanding of the physics of neutron embrittlement should help the life time management of the PWR vessels

Correlation methodology for predicting in-service irradiation embrittlement of reactor pressure vessels

International Nuclear Information System (INIS)

Odette, G.R.

1980-01-01

Irradiation embrittlement of reactor pressure steels is the consequence of altered microstructure due to both irradiation and time-at-temperature. Relatively poor characterisation of the initial microstructure and chemistry, and inaccurate dosimetry and temperature control, as well as failure properly to correlate these variables, have all contributed to a very large scatter in the experimental embrittlement data base. This has made improvement of the basic understanding of embrittlement very difficult. Therefore, it is necessary to develop a more realistic approach to utilising the data base. This is discussed, and proposals are made. (author)

Investigations of low-temperature neutron embrittlement of ferritic steels

International Nuclear Information System (INIS)

Farrell, K.; Mahmood, S.T.; Stoller, R.E.; Mansur, L.K.

1992-01-01

Investigations were made into reasons for accelerated embrittlement of surveillance specimens of ferritic steels irradiated at 50C at the High Flux Isotope Reactor (HFIR) pressure vessel. Major suspects for the precocious embrittlement were a highly thermalized neutron spectrum,a low displacement rate, and the impurities boron and copper. None of these were found guilty. A dosimetry measurement shows that the spectrum at a major surveillance site is not thermalized. A new model of matrix hardening due to point defect clusters indicates little effect of displacement rate at low irradiation temperature. Boron levels are measured at 1 wt ppM or less, inadequate for embrittlement. Copper at 0.3 wt % and nickel at 0.7 wt % are shown to promote radiation strengthening in iron binary alloys irradiated at 50 to 60C, but no dependence on copper and nickel was found in steels with 0.05 to 0.22% Cu and 0.07 to 3.3% Ni. It is argued that copper impurity is not responsible for the accelerated embrittlement of the HFIR surveillance specimens. The dosimetry experiment has revealed the possibility that the fast fluence for the surveillance specimens may be underestimated because the stainless steel monitors in the surveillance packages do not record an unexpected component of neutrons in the spectrum at energies just below their measurement thresholds of 2 to 3 MeV

Nanocrystalline Steels’ Resistance to Hydrogen Embrittlement

Directory of Open Access Journals (Sweden)

Skołek E.

2015-04-01

Full Text Available The aim of this study is to determine the susceptibility to hydrogen embrittlement in X37CrMoV5-1 steel with two different microstructures: a nanocrystalline carbide-free bainite and tempered martensite. The nanobainitic structure was obtained by austempering at the bainitic transformation zone. It was found, that after hydrogen charging, both kinds of microstructure exhibit increased yield strength and strong decrease in ductility. It has been however shown that the resistance to hydrogen embrittlement of X37CrMoV5-1 steel with nanobainitic structure is higher as compared to the tempered martensite. After hydrogen charging the ductility of austempered steel is slightly higher than in case of quenched and tempered (Q&T steel. This effect was interpreted as a result of phase composition formed after different heat treatments.

Embrittling effects of residual elements on steels

International Nuclear Information System (INIS)

Brear, J.M.; King, B.L.

1979-01-01

In a review of work related to reheat cracking in nuclear pressure vessel steels, Dhooge et al referred to work of the authors on the relative embrittling parameter for SA533B steels. The poor agreement when these parameters were applied to creep ductility data for SA508 class 2 lead the reviewers to conclude that the relative importance of impurity elements is a function of base alloy composition. The authors briefly describe some of their more recent work which demonstrates that when various mechanical, and other, effects are taken into consideration, the relative effects of the principal residual elements are similar, despite differing base compositions, and that the embrittling parameters derived correlate well with the data for SA Class 2 steel. (U.K.)

Severe embrittlement of neutron irradiated austenitic steels arising from high void swelling

Energy Technology Data Exchange (ETDEWEB)

Neustroev, V.S. [FSUE ' SSC RF Research Institute of Atomic Reactors' , Dimitrovgrad (Russian Federation)], E-mail: neustroev@niiar.ru; Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

2009-04-30

Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components and introducing limitations on low temperature handling especially. It is shown that the degradation is actually a form of quasi-embrittlement arising from intense flow localization with high levels of localized ductility involving micropore coalescence and void-to-void cracking. Voids initially serve as hardening components whose effect is overwhelmed by the void-induced reduction in shear and Young's moduli at high swelling levels. Thus the alloy appears to soften even as the ductility plunges toward zero on a macroscopic level although a large amount of deformation occurs microscopically at the failure site. Thus the failure is better characterized as 'quasi-embrittlement' which is a suppression of uniform deformation. This case should be differentiated from that of real embrittlement which involves the complete suppression of the material's capability for plastic deformation.

Mercury embrittlement of Cu-Al alloys under cyclic loading

Science.gov (United States)

Regan, T. M.; Stoloff, N. S.

1977-01-01

The effect of mercury on the room temperature, high cycle fatigue properties of three alloys: Cu-5.5 pct Al, Cu-7.3 pct Al, and Cu-6.3 pct Al-2.5 pct Fe has been determined. Severe embrittlement under cyclic loading in mercury is associated with rapid crack propagation in the presence of the liquid metal. A pronounced grain size effect is noted under mercury, while fatigue properties in air are insensitive to grain size. The fatigue results are discussed in relation to theories of adsorption-induced liquid metal embrittlement.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Approach for estimating post-annual reirradiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Server, W.L.; Taboada, A.

1985-01-01

Thermal annealing of a commercial nuclear reactor pressure vessel is a possible solution for extending lifetime in situations where excessive radiation embrittlement has taken place or when the original design life is approached. Two difficult facets of thermal annealing are the degree of toughness recovery after annealing and the post-anneal reirradiation embrittlement behavior. These aspects of annealing are evaluated in this paper by using simple models and translation of the initial irradiation damage curve either vertically or laterally at the point of residual shift after annealing. Results using this methodology are compared to limited actual weld metal measurements of annealing behavior. A forthcoming ASTM Guide on in-place annealing uses this methodology to assess annealing recovery and re-embrittlement response

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Multiscale modelling and experimentation of hydrogen embrittlement in aerospace materials

Science.gov (United States)

Jothi, Sathiskumar

Pulse plated nickel and nickel based superalloys have been used extensively in the Ariane 5 space launcher engines. Large structural Ariane 5 space launcher engine components such as combustion chambers with complex microstructures have usually been manufactured using electrodeposited nickel with advanced pulse plating techniques with smaller parts made of nickel based superalloys joined or welded to the structure to fabricate Ariane 5 space launcher engines. One of the major challenges in manufacturing these space launcher components using newly developed materials is a fundamental understanding of how different materials and microstructures react with hydrogen during welding which can lead to hydrogen induced cracking. The main objective of this research has been to examine and interpret the effects of microstructure on hydrogen diffusion and hydrogen embrittlement in (i) nickel based superalloy 718, (ii) established and (iii) newly developed grades of pulse plated nickel used in the Ariane 5 space launcher engine combustion chamber. Also, the effect of microstructures on hydrogen induced hot and cold cracking and weldability of three different grades of pulse plated nickel were investigated. Multiscale modelling and experimental methods have been used throughout. The effect of microstructure on hydrogen embrittlement was explored using an original multiscale numerical model (exploiting synthetic and real microstructures) and a wide range of material characterization techniques including scanning electron microscopy, 2D and 3D electron back scattering diffraction, in-situ and ex-situ hydrogen charged slow strain rate tests, thermal spectroscopy analysis and the Varestraint weldability test. This research shows that combined multiscale modelling and experimentation is required for a fundamental understanding of microstructural effects in hydrogen embrittlement in these materials. Methods to control the susceptibility to hydrogen induced hot and cold cracking and

Irradiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Bros, J.

2000-01-01

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

Investigation of moisture-induced embrittlement of iron aluminides. Final report

Energy Technology Data Exchange (ETDEWEB)

Alven, D.A.; Stoloff, N.S. [Rensselaer Polytechnic Inst., Troy, NY (United States). Materials Engineering Dept.

1997-06-05

Iron-aluminum alloys with 28 at.% Al and 5 at.% Cr were shown to be susceptible to hydrogen embrittlement by exposure to both gaseous hydrogen and water vapor. This study examined the effect of the addition of zirconium and carbon on the moisture-induced hydrogen embrittlement of an Fe{sub 3}Al,Cr alloy through the evaluation of tensile properties and fatigue crack growth resistance in hydrogen gas and moisture-bearing air. Susceptibility to embrittlement was found to vary with the zirconium content while the carbon addition was found to only affect the fracture toughness. Inherent fatigue crack growth resistance and fracture toughness, as measured in an inert environment, was found to increase with the addition of 0.5 at.% Zr. The combined addition of 0.5 at.% Zr and carbon only increased the fracture toughness. The addition of 1 at.% Zr and carbon was found to have no effect on the crack growth rate when compared to the base alloy. Susceptibility to embrittlement in moisture-bearing environments was found to decrease with the addition of 0.5 at.% Zr. In gaseous hydrogen, the threshold value of the Zr-containing alloys was found to increase above that found in the inert environment while the crack growth resistance was much lower. By varying the frequency of fatigue loading, it was shown that the corrosion fatigue component of the fatigue crack growth rate in an embrittling environment displays a frequency dependence. Hydrogen transport in iron aluminides was shown to occur primarily by a dislocation-assisted transport mechanism. This mechanism, in conjunction with fractography, indicates that the zirconium-containing precipitates act as traps for the hydrogen that is carried along by the dislocations through the lattice.

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)

Status and task of the study on the hydrogen embrittlement of zirconium alloys

International Nuclear Information System (INIS)

Nagase, Fumihisa; Furuta, Teruo; Seino, Shun; Komatsu, Kazushi.

1995-08-01

As the burnup of the LWR fuel is extended, waterside corrosion and hydrogen pickup increase in the Zircaloy cladding. Hydrogen embrittlement of Zircaloy is one of the main factors which may limit the life of the fuel rod. This report presents a review on the hydrogen embrittlement of zirconium and its alloys including the irradiated materials. Research tasks for the reduction of ductility in the high burnup fuel cladding are also discussed. Many fundamental investigations have been performed on the hydrogen embrittlement of zirconium alloys. However, the embrittlement mechanism of the high burnup fuel cladding is complicated. Especially, a coupled effect of hydrides and radiation defects are expected to be pronounced with neutron dose increase. In order to evaluate the reduction of ductility of the higher burnup fuel cladding properly, it is necessary to investigate the coupled effect of these two factors by systematic examinations. (author) 64 refs

Embrittlement of nickel-, cobalt-, and iron-base superalloys by exposure to hydrogen

Science.gov (United States)

Gray, H. R.

1975-01-01

Five nickel-base alloys (Inconel 718, Udimet 700, Rene 41, Hastelloy X, and TD-NiCr), one cobalt-base alloy (L-605), and an iron-base alloy (A-286) were exposed in hydrogen at 0.1 MN/sq m (15 psi) at several temperatures in the range from 430 to 980 C for as long as 1000 hours. These alloys were embrittled to varying degrees by such exposures in hydrogen. Embrittlement was found to be: (1) sensitive to strain rate, (2) reversible, (3) caused by large concentrations of absorbed hydrogen, and (4) not associated with any detectable microstructural changes in the alloys. These observations are consistent with a mechanism of internal reversible hydrogen embrittlement.

Power reactor embrittlement data base (PR-EDB): Uses in evaluating radiation embrittlement of reactor vessels

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1992-01-01

Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current Codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed, computerized data base. Also, such a data is essential for the evaluation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current compilation contains data from 92 reactors and consists of 175 data points for weld materials (79 different welds) and 395 data points for base materials (110 different base materials). The different types of data that are implemented or planned for this data base are discussed. ''User-friendly'' utility programs have been written to investigate a list of problems using this data base. The utility programs are also used to add and upgrade data, retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in this paper

Effect of trapping and temperature on the hydrogen embrittlement susceptibility of alloy 718

Energy Technology Data Exchange (ETDEWEB)

Galliano, Florian; Andrieu, Eric; Blanc, Christine; Cloue, Jean-Marc; Connetable, Damien; Odemer, Gregory, E-mail: gregory.odemer@ensiacet.fr

2014-08-12

Ni-based alloy 718 is widely used to manufacture structural components in the aeronautic and nuclear industries. Numerous studies have shown that alloy 718 may be sensitive to hydrogen embrittlement. In the present study, the susceptibilities of three distinct metallurgical states of alloy 718 to hydrogen embrittlement were investigated to identify both the effect of hydrogen trapping on hydrogen embrittlement and the role of temperature in the hydrogen-trapping mechanism. Cathodic charging in a molten salt bath was used to saturate the different hydrogen traps of each metallurgical state. Tensile tests at different temperatures and different strain rates were carried out to study the effect of hydrogen on mechanical properties and failure modes, in combination with hydrogen content measurements. The results demonstrated that Ni-based superalloy 718 was strongly susceptible to hydrogen embrittlement between 25 °C and 300 °C, and highlighted the dominant roles played by the hydrogen solubility and the hydrogen trapping on mechanical behavior and fracture modes.

Long-term embrittlement of cast duplex stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1990-08-01

This progress report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems during the six months from April to September 1988. Characteristics of the primary mechanism of aging embrittlement (i.e., spinodal decomposition of ferrite) and synergistic effects of alloying and impurity elements that influence the kinetics of the primary mechanism are discussed. Several secondary metallurgical processes of embrittlement, strongly dependent on the C, N, Ni, Mo, and Si content of various heats, are identified. Information on kinetics and data on impact properties are analyzed and correlated with microstructural characteristics to provide a unified method of extrapolating accelerated-aging data to reactor operating conditions. Fracture toughness data are presented for several heats of cast stainless steel aged at temperatures between 320 and 450 degrees C for times up to 10,000 h. Mechanical property data are analyzed to develop the procedure and correlations or predicting the kinetics and extent of embrittlement of reactor components from known material parameters. The method and examples of estimating the impact strength and fracture toughness of cast components during reactor service are described. The lower-bound values of impact strength and fracture toughness for cast stainless steels at LWR operating temperatures are defined. 42 refs., 14 figs., 6 tabs

Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

International Nuclear Information System (INIS)

Krasikov, E.A.; Nikolaenko, V.A.

2008-01-01

Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

International Nuclear Information System (INIS)

2005-06-01

This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

Metal induced embrittlement. Annual report, [March 1, 1987--February 29, 1988

Energy Technology Data Exchange (ETDEWEB)

Hoagland, R.G.

1988-11-01

This program is investigating the causes of embrittlement that occur in certain solid metals when exposed to liquid metals. The degree of embrittlement varies enormously among different solid/liquid pairs as witness, for example, the modest loss of load carrying, ability induced in carbon steels by Pb or the profound embrittlment of aluminum (particularly high strength) alloys by Hg and Ga. The structure of this study involves two types of activities: an experimental fracture mechanics study of the behavior of certain solid metals in liquid metals, and a theoretical study on an atomic scale of the crack tip deformation and extension behavior by means of atomistic simulation. This research, which began March 1, 1987, has completed its 20 month. A brief synopsis is given of performance in each of the areas of activity during the past year.

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

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

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

Influence of sulfur, phosphorus, and antimony segregation on the intergranular hydrogen embrittlement of nickel

International Nuclear Information System (INIS)

Bruemmer, S.M.; Baer, D.R.; Jones, R.H.; Thomas, M.T.

1983-01-01

The effectiveness of sulfur, phosphorus, and antimony in promoting the intergranular embrittlement of nickel was investigated using straining electrode tests in 1N H 2 SO 4 at cathodic potentials. Sulfur was found to be the critical grain boundary segregant due to its large enrichment at grain boundaries (10 4 to 10 5 times the bulk content) and the direct relationship between sulfur coverage and hydrogeninduced intergranular failure. Phosphorus was shown to be significantly less effective than sulfur or antimony in inducing the intergranular hydrogen embrittlement of nickel. The addition of phosphoru to nickel reduced the tendency for intergranular fracture and improved ductility because phosphoru segregated strongly to grain interfaces and limited sulfur enrichment. The hydrogen embrittling potency of antimony was also less than that of sulfur while its segregation propensity was considerably less. It was found that the effectiveness of segregated phosphorus and antimony in prompting inter granular embrittlement vs that of sulfur could be expressed in terms of an equivalent grain boundary sulfur coverage. The relative hydrogen embrittling potencies of sulfur, phosphorus, and antimony are discussed in reference to general mechanisms for the effect of impurity segregation on hydrogeninduced intergranular fracture

Effect of neutron irradiation on vanadium alloys

International Nuclear Information System (INIS)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

Effect of neutron irradiation on vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

Hydrogen embrittlement of Zr-2.5Nb PT with temperature

International Nuclear Information System (INIS)

Oh, Dong Joon; Ahn, Sang Bok; Kim, Young Suk

2003-01-01

The aim of this study is to investigate the effect of hydrogen embrittlement of Zr-2.5Nb CANDU pressure tube. The tests were performed at three hydrogen contents for transverse tensile and CCT specimens while the test temperatures were changed (RT to 300 .deg. C). The specimens were directly machined from the tube retaining original curvature using electric discharge machine. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over 250 .deg. C

Recrystallization and embrittlement of sintered tungsten

International Nuclear Information System (INIS)

Bega, N.D.; Babak, A.V.; Uskov, E.I.

1982-01-01

The recrystallization of sintered tungsten with a cellular structure of deformation is studied as related to its embrittlement. It is stated that in case of preliminary recrystallization the sintered tungsten crack resistance does not depend on the testing temperature. The tungsten crack resistance is shown to lower with an increase of the structure tendency to primary recrystallization [ru

Embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

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

1998-01-01

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

Effect of heat treatments on the hydrogen embrittlement ...

Indian Academy of Sciences (India)

pipe steel in as received (controlled rolled), normalized, and quenched and tempered conditions. The resistance to hydrogen embrittlement was found in the order of controlled rolled > quenched and tempered > normalized. The fracture mode ...

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments

International Nuclear Information System (INIS)

Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

1984-01-01

In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700 0 C, but the creep tests at 800 0 C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X

Oak Ridge National Laboratory Embrittlement Data Base (EDB) and Dosimetry Evaluation (DE) program

International Nuclear Information System (INIS)

Pace, J.V. III; Remec, I.; Wang, J.A.; White, J.E.

1996-01-01

The objective of this program is to develop, maintain, and upgrade computerized data bases, calculational procedures, and standards relating to reactor pressure vessel fluence spectra determinations and embrittlement assessments. As part of this program, the information from radiation embrittlement research on nuclear reactor pressure vessel steels and from power reactor surveillance reports is maintained in a data base published on a periodic basis. The Embrittlement Data Base (EDB) effort consists of verifying the quality of the EDB, providing user-friendly software to access and process the data, and exploring and assessing embrittlement prediction models. The Dosimetry Evaluation effort consists of maintaining and upgrading validated neutron and gamma radiation transport procedures, maintaining cross-section libraries with the latest evaluated nuclear data, and maintaining and updating validated dosimetry procedures and data bases. The information available from this program provides data for assisting the Office of Nuclear Reactor Regulation, with support from the Office of Nuclear Regulatory Research, to effectively monitor current procedures and data bases used by vendors, utilities, and service laboratories in the pressure vessel irradiation surveillance program

Evaluation of temper embrittlement of martensitic and ferritic-martensitic steels by acoustic emission

International Nuclear Information System (INIS)

Lu, Yusho; Takahashi, Hideaki; Shoji, Tetsuo

1987-01-01

Martensitic (HT-9) and ferritic-martensitic steels (9Cr-2Mo) are considered as fusion first wall materials. In this investigation in order to understand the sensitivity of temper embrittlement in these steels under actual service condition, fracture toughness testing was made by use of acoustic emission technique. The temper embrittlement was characterized in terms of fracture toughness. The fracture toughness of these steels under 500 deg C, 100 hrs, and 1000 hrs heat treatment was decreased and their changes in micro-fracture process have been observed. The fracture toughness changes by temper embrittlement was discussed by the characteristic of AE, AE spectrum analysis and fractographic investigation. The relation between micro-fracture processes and AE has been clarified. (author)

Multiscale Modeling of Hydrogen Embrittlement for Multiphase Material

KAUST Repository

Al-Jabr, Khalid A.

2014-01-01

Hydrogen Embrittlement (HE) is a very common failure mechanism induced crack propagation in materials that are utilized in oil and gas industry structural components and equipment. Considering the prediction of HE behavior, which is suggested

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

International Nuclear Information System (INIS)

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

1999-03-01

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

Hydrogen environment embrittlement

International Nuclear Information System (INIS)

Donovan, J.A.

1975-01-01

Exposure of many metals to gaseous hydrogen causes losses in elongation, reduction of area, and fracture toughness, and causes increases in slow crack growth rate or fatigue life compared with values obtained in air or vacuum. Hydrogen pressure, temperature, and purity significantly influence deleterious effects. The strength and structural characteristics of the metal influence the degradation of its properties by hydrogen. Several theories have been proposed to explain the loss of properties in hydrogen, but none has gained wide acceptance. The embrittlement mechanism and the role of diffusion are, therefore, open questions and need more quantitative experimental data both to test the proposed theories and to allow the development of realistic preventive measures. (U.S.)

The influences of impurity content, tensile strength, and grain size on in-service temper embrittlement of CrMoV steels

International Nuclear Information System (INIS)

Cheruvu, N.S.; Seth, B.B.

1989-01-01

The influences of impurity levels, grain size, and tensile strength on in-service temper embrittlement of CrMoV steels have been investigated. The samples for this study were taken from steam turbine CrMoV rotors which had operated for 15 to 26 years. The effects of grain size and tensile strength on embrittlement susceptibility were separated by evaluating the embrittlement behavior of two rotor forgings made from the same ingot after an extended step-cooling treatment. Among the residual elements in the steels, only P produces a significant embrittlement. The variation of P and tensile strength has no effect on in-service temper embrittlement susceptibility, as measured by the shift in fracture appearance transition temperature (FATT). However, the prior austenite grain size plays a major role in service embrittlement. The fine grain steels with a grain size of ASTM No. 9 or higher are virtually immune to in-service embrittlement. In steels having duplex grain sizes, embrittlement susceptibility is controlled by the size of coarser grains. For a given steel chemistry, the coarse grain steel is more susceptible to in-service embrittlement, and a decrease in ASTM grain size number from 4 to 0/1 increases the shift in FATT by 61 degrees C (10/10 degrees F). It is demonstrated that long-term service embrittlement can be simulated, except in very coarse grain steels, by using the extended step-cooling treatment. The results of step-cooling studies show that the coarse grain rotor steels take longer time during service to reach a fully embrittled state than the fine grain rotor steels

Hydrogen embrittlement of titanium tested with fracture mechanics specimens

International Nuclear Information System (INIS)

Aho-Mantila, I.; Rahko, P.

1990-11-01

Titanium is one of the possible canister materials for spent nuclear fuel. The aim of this study is to determine whether the hydrogen embrittlement of titanium could be a possible deterioration mechanism of titanium canisters. This experimental study was preceded by a literature review and an experimental study on crack nucleation. Tests in this study were carried out with hydrogen charged fracture mechanics specimens. The studied hydrogen contents were as received, 100 ppm, 200 ppm, 500 ppm and 700 ppm and the types of the studied titanium were ASTM Grades 2 and 12. Test methods were slow tensile test (0.027 mm/h) and fatigue test (stress ratio 0.7 or 0.8 and frequency 5 Hz). According to the literature titanium may be embrittled by hydrogen at slow strain rates and cracking may occur under sustained load. In this study no evidence of hydrogen embrittlement was noticed in slow strain rate tension with bulk hydrogen contents up to 700 ppm. The fatigue tests of titanium Grades 2 and 12 containing 700 ppm hydrogen showed even slower crack growth compared to the as received condition. Very high hydrogen contents well in eccess of 700 ppm on the surface of titanium can, however, facilitate surface crack nucleation and crack growth, as shown in the previous study

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

International Nuclear Information System (INIS)

Sudo, Akira; Miyazono, Shohachiro

1983-06-01

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

Role of twinning and transformation in hydrogen embrittlement of austenitic stainless steels

International Nuclear Information System (INIS)

Caskey, G.R. Jr.

1977-01-01

Internal hydrogen embrittlement may be viewed as an extreme form of environmental embrittlement that arises following prolonged exposure to a source of hydrogen. Smooth bar tensile specimens of three stainless steels saturated with deuterium (approximately 200 mol D 2 /m 3 ) were pulled to failure in air at 200 to 400 0 K or in liquid nitrogen at 78 0 K. In Type 304L stainless steel and Tenelon ductility losses are a maximum around 200 to 273 0 K; Type 310 stainless steel is not embrittled at this hydrogen concentration. A distinct change in fracture mode accompanies hydrogen embrittlement, with fracture proceeding along coherent boundaries of pre-existing annealing twins. This fracture path is observed in Tenelon at 78 0 K even when hydrogen is absent. There is also a change in fracture appearance in specimens with no prior exposure to hydrogen if they are pulled to failure in high-pressure hydrogen. The fracture path is not identifiable, however. Magnetic response measurements and changes in the stress-strain curves show that hydrogen suppresses formation of strain-induced α'-martensite at 198 0 K in both Type 304L stainless steel and Tenelon, but there is little effect in Type 304L stainless at 273 0 K

Ni/boride interfaces and environmental embrittlement in Ni-based superalloys: A first-principles study

International Nuclear Information System (INIS)

Sanyal, Suchismita; Waghmare, Umesh V.; Hanlon, Timothy; Hall, Ernest L.

2011-01-01

Highlights: ► Fracture strengths of Ni/boride interfaces through first-principles calculations. ► Fracture strengths of Ni/boride interfaces are higher than Ni/Ni 3 Al and NiΣ5 grain boundaries. ► Ni/boride interfaces have higher resistance to O-embrittlement than Ni/Ni 3 Al and NiΣ5 grain boundaries. ► CrMo-borides are more effective than Cr-borides in resisting O-embrittlement. ► Electronegativity differences between alloying elements correlate with fracture strengths. - Abstract: Motivated by the vital role played by boride precipitates in Ni-based superalloys in improving mechanical properties such as creep rupture strength, fatigue crack growth rates and improved resistance towards environmental embrittlement , we estimate fracture strength of Ni/boride interfaces through determination of their work of separation using first-principles simulations. We find that the fracture strength of Ni/boride interfaces is higher than that of other commonly occurring interfaces in Ni-alloys, such as Ni Σ-5 grain boundaries and coherent Ni/Ni 3 Al interfaces, and is less susceptible to oxygen-induced embrittlement. Our calculations show how the presence of Mo in Ni/M 5 B 3 (M = Cr, Mo) interfaces leads to additional reduction in oxygen-induced embrittlement. Through Electron-Localization-Function based analyses, we identify the electronic origins of effects of alloying elements on fracture strengths of these interfaces and observe that chemical interactions stemming from electronegativity differences between different atomic species are responsible for the trends in calculated strengths. Our findings should be useful towards designing Ni-based alloys with higher interfacial strengths and reduced oxygen-induced embrittlement.

Charles J. McMahon Interfacial Segregation and Embrittlement Symposium

National Research Council Canada - National Science Library

Vitek, Vaclav

2003-01-01

.... McMahon Interfacial Segregation and Embrittlement Symposium: Grain Boundary Segregation and Fracture in Steels was sponsored by ASM International, Materials Science Critical Technology Sector, Structural Materials Division, Materials Processing...

Role of vanadium carbide traps in reducing the hydrogen embrittlement susceptibility of high strength alloy steels. Final report

Energy Technology Data Exchange (ETDEWEB)

Spencer, G.L.; Duquette, D.J.

1998-08-01

High strength alloy steels typically used for gun steel were investigated to determine their susceptibility to hydrogen embrittlement. Although AISI grade 4340 was quite susceptible to hydrogen embrittlement, ASTM A723 steel, which has identical mechanical properties but slightly different chemistries, was not susceptible to hydrogen embrittlement when exposed to the same conditions. The degree of embrittlement was determined by conducting notched tensile testing on uncharged and cathodically charged specimens. Chemical composition was modified to isolate the effect of alloying elements on hydrogen embrittlement susceptibility. Two steels-Modified A723 (C increased from 0.32% to 0.40%) and Modified 4340 (V increased from 0 to O.12%) were tested. X-ray diffraction identified the presence of vanadium carbide, V{sub 4}C{sub 3}, in A-23 steels, and subsequent hydrogen extraction studies evaluated the trapping effect of vanadium carbide. Based on these tests, it was determined that adding vanadium carbide to 4340 significantly decreased hydrogen embrittlement susceptibility because vanadium carbide traps ties up diffusible hydrogen. The effectiveness of these traps is examined and discussed in this paper.

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Najzer, M.

1982-01-01

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

Present status of the disk pressure tests for hydrogen embrittlement

International Nuclear Information System (INIS)

Fidelle, J.P.

1985-05-01

The Disk Pressure Tests (DPT) have been developed considerably theoretically and experimentally for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for Environment Embrittlement due to H 2 , hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably for pressure up to 300 MPa and temperature (-160 0 C to 1000 0 C). Very low strain rate -longer than a month- tests have been able to evidence embrittlement of FFC alloys where H diffusivity is low. Conversely for very oxidation - sensitive metals (e.g. Nb and Ta) effects may appear only at somewhat high rates. The relationship between dynamic (increasing stress) tests, static (delayed failure) and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analyzed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 150 + materials in different conditions. From the tests on a large number of metal systems, a theory of HE has been derived which accounts for the behavior of metals and alloys either embrittled and or hydrited. Finally comparison of HGE tests and service behavior of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service

Effect of microstructure on the susceptibility of a 533 steel to temper embrittlement

International Nuclear Information System (INIS)

Raoul, S.; Marini, B.; Pineau, A.

1998-01-01

In ferritic steels, brittle fracture usually occurs at low temperature by cleavage. However the segregation of impurities (P, As, Sn etc..) along prior γ grain boundaries can change the brittle fracture mode from transgranular to intergranular. In quenched and tempered steels, this segregation is associated with what is called the temper-embrittlement phenomenon. The main objective of the present study is to investigate the influence of the as-quenched microstructure (lower bainite or martensite) on the susceptibility of a low alloy steel (A533 cl.1) to temper-embrittlement. Dilatometric tests were performed to determine the continous-cooling-transformation (CCT) diagram of the material and to measure the critical cooling rate (V c ) for a martensitic quench. Then subsized Charpy V-notched specimens were given various cooling rates from the austenitization temperature to obtain a wide range of as-quenched microstructures, including martensite and bainite. These specimens were subsequently given a heat treatment to develop temper embrittlement and tested to measure the V-notch fracture toughness at -50 C. The fracture surfaces were examined by SEM. It is shown that martensitic microstructures are more susceptible to intergranular embrittlement than bainitic microstructures. These observed microstructural influences are briefly discussed. (orig.)

On the tempered martensite embrittlement in AISI 4140 low alloy steel

Energy Technology Data Exchange (ETDEWEB)

Darwish, F.A. (Dept. of Materials Science and Metallurgy, Catholic Univ., Rio de Janeiro, RJ (Brazil)); Pereira, L.C.; Gatts, C. (Dept. of Metallurgy and Materials Engineering, Federal Univ., Rio de Janeiro, RJ (Brazil)); Graca, M.L. (Materials Div., Technical Aerospace Center, Sao Jose dos Campos, SP (Brazil))

1991-02-01

In the present investigation the Auger electron spectroscopy (AES) technique was used to determine local carbon and phosphorus concentrations on the fracture surfaces of as-quenched and quenched-and-tempered (at 350deg C) AISI 4140 steel specimens austenitized at low and high temperatures. The AES results were rationalized to conclude that, although carbide growth as well as phosphorus segregation are expected to contribute to tempered martensite embrittlement, carbide precipitation on prior austenite grain boundaries during tempering is seen to be the microstructural change directly responsible for the occurrence of the referred embrittlement phenomenon. (orig.).

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

Precipitation hardening and hydrogen embrittlement of aluminum ...

Indian Academy of Sciences (India)

Hydrogen susceptibility of alloy AA7020 was evaluated by slow strain-rate tensile ... high pressures because of the embrittling effect of hydrogen. ... The higher the total Zn + Mg content,. ∗ .... dislocations, leading to a local softening of the slip plane, and thus to ... A Vickers hardness testing machine was used to measure the.

Low temperature thermal ageing embrittlement of austenitic stainless steel welds and its electrochemical assessment

International Nuclear Information System (INIS)

Chandra, K.; Kain, Vivekanand; Raja, V.S.; Tewari, R.; Dey, G.K.

2012-01-01

Highlights: ► Embrittlement study of austenitic stainless steel welds after ageing up to 20,000 h. ► Spinodal decomposition and G-phase precipitation in ferrite at 400 °C. ► Spinodal decomposition of ferrite at 335 and 365 °C. ► Large decrease in corrosion resistance due to G-phase precipitation. ► Good correlation between electrochemical properties and the degree of embrittlement. - Abstract: The low temperature thermal ageing embrittlement of austenitic stainless steel welds is investigated after ageing up to 20,000 h at 335, 365 and 400 °C. Spinodal decomposition and G-phase precipitation after thermal ageing were identified by transmission electron microscopy. Ageing led to increase in hardness of the ferrite phase while there was no change in the hardness of austenite. The degree of embrittlement was evaluated by non-destructive methods, e.g., double-loop and single-loop electrochemical potentiokinetic reactivation tests. A good correlation was obtained between the electrochemical properties and hardening of the ferrite phase of the aged materials.

Effect of hydrogen and oxygen content on the embrittlement of Zr alloys

International Nuclear Information System (INIS)

Griger, A.; Hozer, Z.; Matus, L.; Vasaros, L.; Horvath, M.

2001-01-01

An experimental study is carried out in the KFKI Atomic Energy Research Institute in order to clear up the role of oxidation and hydrogen uptake in the embrittlement process. Russian E110 type Zr1%Nb and Zircaloy-4 claddings are used as test materials. The differences between the properties of two alloys are examined. The sample preparation covered the following cases: oxidation in Ar+O 2 atmosphere; hydrogen uptake of as received and pre-oxidised samples (in Ar+O 2 atmosphere); oxidation in steam. The oxidation in Ar+O 2 and the subsequent hydrogen uptake procedure make possible the production of samples with well-characterized hydrogen and oxygen content. Corrosion treated ring samples of 8 mm height are examined in ring compression tests. The force-deformation curves are recorded and the crushing force and deformation are determined. The relative deformation is used for the characterisation of embrittlement level. The results of experiments provide detailed information about the effect of hydrogen and oxygen content on the embrittlement of zirconium alloys. The conclusions are: 1) hydrogen seems to play a more important role in the embrittlement of zirconium alloys than oxygen; 2) the Zircaloy-4 alloy becomes brittle at lower hydrogen content than the Zr1%Nb; 3) under steam oxidation conditions the Zr1%Nb alloy takes up much more hydrogen and becomes more brittle than the Zircaloy-4

Comparison of hydrogen gas embrittlement of austenitic and ferritic stainless steels

Science.gov (United States)

Perng, T. P.; Altstetter, C. J.

1987-01-01

Hydrogen-induced slow crack growth (SCG) was compared in austenitic and ferritic stainless steels at 0 to 125 °Cand 11 to 216 kPa of hydrogen gas. No SCG was observed for AISI 310, while AISI 301 was more susceptible to hydrogen embrittlement and had higher cracking velocity than AL 29-4-2 under the same test conditions. The kinetics of crack propagation was modeled in terms of the hydrogen transport in these alloys. This is a function of temperature, microstructure, and stress state in the embrittlement region. The relatively high cracking velocity of AISI 301 was shown to be controlled by the fast transport of hydrogen through the stress-induced α' martensite at the crack tip and low escape rate of hydrogen through the γ phase in the surrounding region. Faster accumulation rates of hydrogen in the embrittlement region were expected for AISI 301, which led to higher cracking velocities. The mechanism of hydrogen-induced SCG was discussed based upon the concept of hydrogen-enhanced plasticity.

Heavy-Section Steel Irradiation Program: Embrittlement issues

International Nuclear Information System (INIS)

Corwin, W.R.

1991-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. It is imperative to understand and predict the capabilities and limitations of its integrity. It is particularly vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. The Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Results from HSSI studies provide information needed to aid in resolving major regulatory issues facing the USNRC which involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf (LUS) welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. Embrittlement modeling studies have shown that the time or dose required for the point defect concentrations, which ultimately contribute to irradiation embrittlement, to reach their steady state values can be comparable to the component lifetime or to the duration of an irradiation experiment

Current limitations of trend curve analysis for the prediction of reactor PV embrittlement

International Nuclear Information System (INIS)

Gold, R.; McElroy, W.N.

1986-02-01

In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration since the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant

Study of intergranular embrittlement in Fe-12Mn alloys

International Nuclear Information System (INIS)

Lee, H.J.

1982-06-01

A high resolution scanning Auger microscopic study has been performed on the intergranular fracture surfaces of Fe-12Mn steels in the as-austenitized condition. Fracture mode below the ductile-brittle transition temperature was intergranular whenever the alloy was quenched from the austenite field. The intergranular fracture surface failed to reveal any consistent segregation of P, S, As, O, or N. The occasional appearance of S or O on the fracture surface was found to be due to a low density precipitation of MnS and MnO 2 along the prior austenite boundaries. An AES study with Ar + ion-sputtering showed no evidence of manganese enrichment along the prior austenite boundaries, but a slight segregation of carbon which does not appear to be implicated in the tendency toward intergranular fracture. Addition of 0.002% B with a 1000 0 C/1h/WQ treatment yielded a high Charpy impact energy at liquid nitrogen temperature, preventing the intergranular fracture. High resolution AES studies showed that 3 at. % B on the prior austenite grain boundaries is most effective in increasing the grain boundary cohesive strength in an Fe-12Mn alloy. Trace additions of Mg, Zr, or V had negligible effects on the intergranular embrittlement. A 450 0 C temper of the boron-modified alloys was found to cause tempered martensite embrittlement, leading to intergranular fracture. The embrittling treatment of the Fe-12Mn alloys with and without boron additions raised the ductile-brittle transition by 150 0 C. This tempered martensite embrittlement was found to be due to the Mn enrichment of the fracture surface to 32 at. % Mn in the boron-modified alloy and 38 at. % Mn in the unmodified alloy. The Mn-enriched region along the prior austenite grain boundaries upon further tempering is believed to cause nucleation of austenite and to change the chemistry of the intergranular fracture surfaces. 61 figures

Experimental study on the resistance to hydrogen embrittlement of NIFS-V4Cr4Ti alloy

International Nuclear Information System (INIS)

Chen Jiming; Xu Zengyu; Den Ying; Muroga, T.

2002-01-01

SWIP (Southwestern Institute of Physics) has joined an international collaboration on the hydrogen embrittlement resistance evaluation of the vanadium alloy. This paper presents some experiments on the tensile properties and Charpy impact properties of the NIFS-V4Cr4Ti alloy with high-level hydrogen concentration. The experiment results show different properties against hydrogen embrittlement in static tension and impact load. The critical hydrogen concentration required to embrittle the alloy was about 215 - 310 mg·kg -1 on static tension load, but less than 130 mg·kg -1 on impact loading

Irreversible traps, their influence on the embrittlement of high strength steel

International Nuclear Information System (INIS)

Mariano, I; Mansilla, G

2012-01-01

Hydrogen (H) can be trapped in lattice defects such as vacancies, dislocations, grain boundaries and interfaces between the matrix and precipitates. The effect on the mechanical properties depends on factors inherent in materials such as the activation energy of irreversible traps (H trapped in Network Places) and its sensitivity to embrittlement. Differential scanning calorimetry (DSC) allows the study of those processes in which enthalpy variation occurs. The purpose is to record the difference in enthalpy change that occurs in the sample as a function of temperature or time. This work represents a study of H embrittlement of high strength steel resulfurized

Effect of microstructure on the susceptibility of a 533 steel to temper embrittlement

Energy Technology Data Exchange (ETDEWEB)

Raoul, S.; Marini, B. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Pineau, A. [CNRS, Evry (France). Centre de Materiaux

1998-11-01

In ferritic steels, brittle fracture usually occurs at low temperature by cleavage. However the segregation of impurities (P, As, Sn etc..) along prior {gamma} grain boundaries can change the brittle fracture mode from transgranular to intergranular. In quenched and tempered steels, this segregation is associated with what is called the temper-embrittlement phenomenon. The main objective of the present study is to investigate the influence of the as-quenched microstructure (lower bainite or martensite) on the susceptibility of a low alloy steel (A533 cl.1) to temper-embrittlement. Dilatometric tests were performed to determine the continous-cooling-transformation (CCT) diagram of the material and to measure the critical cooling rate (V{sub c}) for a martensitic quench. Then subsized Charpy V-notched specimens were given various cooling rates from the austenitization temperature to obtain a wide range of as-quenched microstructures, including martensite and bainite. These specimens were subsequently given a heat treatment to develop temper embrittlement and tested to measure the V-notch fracture toughness at -50 C. The fracture surfaces were examined by SEM. It is shown that martensitic microstructures are more susceptible to intergranular embrittlement than bainitic microstructures. These observed microstructural influences are briefly discussed. (orig.) 11 refs.

Stress corrosion cracking and hydrogen embrittlement of an Al-Zn-Mg-Cu alloy

International Nuclear Information System (INIS)

Song, R.G.; Dietzel, W.; Zhang, B.J.; Liu, W.J.; Tseng, M.K.; Atrens, A.

2004-01-01

The age hardening, stress corrosion cracking (SCC) and hydrogen embrittlement (HE) of an Al-Zn-Mg-Cu 7175 alloy were investigated experimentally. There were two peak-aged states during ageing. For ageing at 413 K, the strength of the second peak-aged state was slightly higher than that of the first one, whereas the SCC susceptibility was lower, indicating that it is possible to heat treat 7175 to high strength and simultaneously to have high SCC resistance. The SCC susceptibility increased with increasing Mg segregation at the grain boundaries. Hydrogen embrittlement (HE) increased with increased hydrogen charging and decreased with increasing ageing time for the same hydrogen charging conditions. Computer simulations were carried out of (a) the Mg grain boundary segregation using the embedded atom method and (b) the effect of Mg and H segregation on the grain boundary strength using a quasi-chemical approach. The simulations showed that (a) Mg grain boundary segregation in Al-Zn-Mg-Cu alloys is spontaneous, (b) Mg segregation decreases the grain boundary strength, and (c) H embrittles the grain boundary more seriously than does Mg. Therefore, the SCC mechanism of Al-Zn-Mg-Cu alloys is attributed to the combination of HE and Mg segregation induced grain boundary embrittlement

Fracture toughness prediction for RPV Steels with various degree of embrittlement

International Nuclear Information System (INIS)

Margolin, B.; Gulenko, A.; Shvetsova, V.

2003-01-01

In the present report, predictions of the temperature dependence of cleavage fracture toughness are performed on the basis of the Master Curve approach and a probabilistic model named now the Prometey model. These predictions are performed for reactor pressure vessel steels in different states, the initial (as-produced), irradiated state with moderate degree of embrittlement and in the highly embrittled state. Calculations of the K IC (T) curves may be performed with both approaches on the basis of fracture toughness test results from pre-cracked Charpy specimens at some (one) temperature. The calculated curves are compared with test results. It is shown that the K IC (T) curves for the initial state calculated with the Master Curve approach and the probabilistic model show good agreement. At the same time, for highly embrittled RPV steel, the K IC (T) curve predicted with the Master Curve approach is not an adequate fit to the experimental data, whereas the agreement of the test results and the K IC (T) curve calculated with the probabilistic model is good. An analysis is performed for a possible variation of the K IC (T) curve shape and the scatter in K IC results. (author)

Effects of metallurgical variables on hydrgen embrittlement in types 316, 321, and 347 stainless steels

International Nuclear Information System (INIS)

Rozenak, P.; Eliezer, D.

1984-01-01

Hydrogen embrittlement of 316, 321 and 347 types austenitic stainless steels has been studied by charging thin tensile specimens with hydrogen through cathodic polarization. Throughout this study we have compared solution annealed samples having various prior austenitic grain-size with samples given the additional sensitization treatment. The results show that refined grains improves the resistance to hydrogen cracking regardless of the failure mode. The sensitized specimens were predominantly intergranular, while the annealed specimens show massive regions of microvoid coalescence producing ductile rupture. 347 type stainless steel is much more susceptible to hydrogen embrittlement than 321 type steel, and 316 type is the most resistant to hydrogen embrittlement. the practical implication of the experimental conclusions are discussed

Zinc-induced embrittlement in nickel-base superalloys by simulation and experiment

Science.gov (United States)

Otis, Richard; Waje, Mahesh; Lindwall, Greta; Jefferson, Tiffany; Lange, Jeremy; Liu, Zi-Kui

2017-09-01

The high cost of Re has driven interest in processes for recovering Re from scrap superalloy parts. In this work thermodynamic modelling is used to study Zn-induced embrittlement of a superalloy and to direct experiments. Treating superalloy powder with Zn vapour reduces the average particle size after milling from approximately ?m to 0.5-10 ?m, vs. ?m for untreated powder. Simulations predict the required treatment time to increase with temperature. Agreement between predictions and experiments suggests that an embrittling liquid forms in less than an hour of Zn vapour treatment between 950-1000 ?C and partial pressures of Zn between 14-34 kPa (2-5 psi).

Initial assessment of the mechanisms and significance of low-temperature embrittlement of cast stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Sather, A.

1990-08-01

This report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems. Metallurgical characterization and mechanical property data from Charpy-impact, tensile, and J-R curve tests are presented for several experimental and commercial heats, as well as for reactor-aged CF-3, CF-8, and CF-8M cast stainless steels. The effects of material variables on the embrittlement of cast stainless steels are evaluated. Chemical composition and ferrite morphology strongly affect the extent and kinetics of embrittlement. In general, the low-carbon CF-3 stainless steels are the most resistant and the molybdenum-containing high-carbon CF-8M stainless steels are most susceptible to embrittlement. The microstructural and mechanical-property data are analyzed to establish the mechanisms of embrittlement. The procedure and correlations for predicting the impact strength and fracture toughness of cast components during reactor service are described. The lower bound values of impact strength and fracture toughness for low-temperature-aged cast stainless steel are defined. 39 refs., 56 figs., 8 tabs

Long-term aging embrittlement of cast duplex stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1991-01-01

The primary objectives of this program are to investigate the significance of in-service embrittlement of cast duplex stainless steels in light water reactor (LWR) systems and to evaluate possible remedies for the embrittlement problem in existing and future plants. The scope of the investigation includes three goals: (1) develop a methodology and correlations for predicting the toughness loss suffered by cast stainless steel components during normal and extended life of LWRs, (2) validate the simulation of in-reactor degradation by accelerated aging, and (3) establish the effects of key compositional and metallurgical variables on the kinetics and extent of embrittlement. The emphasis during the current year was on developing a procedure and correlations for predicting fracture toughness J-R curves of aged cast stainless steels from known material information. The present analysis has focused on developing correlations for the fracture properties in terms of material information that can be determined from the certified material test record (CMTR) and on ensuring that the correlations are adequately conservative for structurally weak materials

Status of reactor pressure vessel embrittlement study in Japan

International Nuclear Information System (INIS)

Sasajima, H.

1997-01-01

Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

Embrittlement of the Shippingport reactor shield tank

International Nuclear Information System (INIS)

Chopra, O.K.; Shack, W.J.

1989-01-01

Surveillance specimens from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory showed an unexpectedly high degree of embrittlement relative to the data obtained on similar materials in Materials Testing Reactors (MTRs). The results suggest a possible negative flux effect and raise the issue of embrittlement of the pressure vessel support structures of commercial light water reactors. To help resolve this issues, a program was initiated to characterize the irradiation embrittlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport NST operated at 55 degree C (130 degree F) and was fabricated from A212 Grade B steel, similar to the vessel material in HFIR. The inner wall of the NST was exposed to a total maximum fluence of ∼ 6 x 10 17 n/cm 2 (E > 1 MeV) over a life of 9.25 effective full power years. This corresponds to a fast flux of 2.1 x 10 9 n/cm 2 x s and is comparable to the conditions for the HFIR surveillance specimens. The results indicate that irradiation increases the 15 ft x lb Charpy transition temperature (CTT) by ∼25 degree C (45 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens and is consistent with that expected from the MTR data base. However, the actual value of CTT is high, and the toughness at service temperature is low, even when compared with the HFIR data. The increase in yield stress is ∼50 MPa, which is comparable to the HFIR data. The results also indicate a lower impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall

Environmental embrittlement of intermetallic compounds in Fe-Al alloys

Institute of Scientific and Technical Information of China (English)

张建民; 张瑞林; S.H.YU; 余瑞璜

1996-01-01

First,it is proposed that hydrogen atoms occupy the interstitial sites in Fe3Al and FeAl.Then the environmental embrittlement of intermetallic compounds in Fe-Al alloys is studied in the light of calculated valence electron structures and bond energy of Fe3Al and FeAl containing hydrogen atoms.From the analyses it is found that the states of metal atoms will change,in which more lattice electrons will become covalent electrons to bond with hydrogen atoms when the atomic hydrogen diffuses into the intermetallic compounds in Fe-Al alloys,which will result in the decrease of local metallicity in Fe3Al and FeAl.Meanwhile,it is found that the crystal will easily cleave since solute hydrogen bonds with metal atoms and severely anisotropic bonds form.As a conclusion,these factors result in the environmental embrittlement of Fe3Al and FeAl.

Oxidation-induced embrittlement and structural changes of Zircaloy-4 tubing in steam at 700-1000 deg. C

Energy Technology Data Exchange (ETDEWEB)

Ali, A E; Huessein, A G; El-Sayed, A A; El Banna, O A [Atomic Energy Authority, Cairo (Egypt); El Raghy, S M [Cairo Univ. (Egypt). Faculty of Engineering

1997-02-01

The oxidation-induced embrittlement and structural changes of Zircaloy-4 (KWU-Type) tubing was investigated under light water reactors (LWR) Loss-of-Coolant. Accident conditions (LOCA) in temperature range 700-1000 deg. C. The effect of hydrogen addition to steam was also investigated in the temperature range 800-1000 deg. C. The oxidation-induced embrittlement was found to be a function of both temperature and time. Fractography investigation of oxidized tubing showed a typical brittle fracture in the stabilized-alpha zone. The microhardness measurements revealed that the alpha-Zr is harder than that near the mid-wall position. The oxidation-induced embrittlement at 900 deg. C was found to be higher than at 1000 deg. C. The results also indicated that the addition of 5% by volume hydrogen to steam resulted in an increase in the degree of embrittlement. (author). 22 refs, 9 figs, 3 tabs.

PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

International Nuclear Information System (INIS)

Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

1990-06-01

Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

Application of magnetomechanical hysteresis modeling to magnetic techniques for monitoring neutron embrittlement and biaxial stress

International Nuclear Information System (INIS)

Sablik, M.J.; Kwun, H.; Rollwitz, W.L.; Cadena, D.

1992-01-01

The objective is to investigate experimentally and theoretically the effects of neutron embrittlement and biaxial stress on magnetic properties in steels, using various magnetic measurement techniques. Interaction between experiment and modeling should suggest efficient magnetic measurement procedures for determining neutron embrittlement biaxial stress. This should ultimately assist in safety monitoring of nuclear power plants and of gas and oil pipelines. In the first six months of this first year study, magnetic measurements were made on steel surveillance specimens from the Indian Point 2 and D.C. Cook 2 reactors. The specimens previously had been characterized by Charpy tests after specified neutron fluences. Measurements now included: (1) hysteresis loop measurement of coercive force, permeability and remanence, (2) Barkhausen noise amplitude; and (3) higher order nonlinear harmonic analysis of a 1 Hz magnetic excitation. Very good correlation of magnetic parameters with fluence and embrittlement was found for specimens from the Indian Point 2 reactor. The D.C. Cook 2 specimens, however showed poor correlation. Possible contributing factors to this are: (1) metallurgical differences between D.C. Cook 2 and Indian Point 2 specimens; (2) statistical variations in embrittlement parameters for individual samples away from the stated men values; and (3) conversion of the D.C. Cook 2 reactor to a low leakage core configuration in the middle of the period of surveillance. Modeling using a magnetomechanical hysteresis model has begun. The modeling will first focus on why Barkhausen noise and nonlinear harmonic amplitudes appear to be better indicators of embrittlement than the hysteresis loop parameters

Cryogenic filter method produces super-pure helium and helium isotopes

Science.gov (United States)

Hildebrandt, A. F.

1964-01-01

Helium is purified when cooled in a low pressure environment until it becomes superfluid. The liquid helium is then filtered through iron oxide particles. Heating, cooling and filtering processes continue until the purified liquid helium is heated to a gas.

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)

Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Takamizawa, Hisashi, E-mail: takamizawa.hisashi@jaea.go.jp; Itoh, Hiroto, E-mail: ito.hiroto@jaea.go.jp; Nishiyama, Yutaka, E-mail: nishiyama.yutaka93@jaea.go.jp

2016-10-15

In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

Cooperation modes of the radiation embrittlement

International Nuclear Information System (INIS)

Voevodin, V.N.; Laptev, I.N.; Neklyudov, I.M.; Ozhigov, L.S.; Bryk, V.V.; Parkhomenko, A.A.

2012-01-01

According to the results of experimental and theoretical studies of the structures and properties of irradiated deformed materials with different crystalline structure, the effect of irradiation on mechanisms of radiation embrittlement on all structure levels (from atomic to macrolevel) has been shown. The effects of structural localization, collectivization, long range effects, rotation modes development are described. It was shown that these effects are closely interrelated; they characterized the deformed irradiation material as open dissipative system subjected to the laws of such scientific approach as synergetic.

Hydrogen Embrittlement Mechanism in Fatigue Behavior of Austenitic and Martensitic Stainless Steels

Directory of Open Access Journals (Sweden)

Sven Brück

2018-05-01

Full Text Available In the present study, the influence of hydrogen on the fatigue behavior of the high strength martensitic stainless steel X3CrNiMo13-4 and the metastable austenitic stainless steels X2Crni19-11 with various nickel contents was examined in the low and high cycle fatigue regime. The focus of the investigations were the changes in the mechanisms of short crack propagation. Experiments in laboratory air with uncharged and precharged specimen and uncharged specimen in pressurized hydrogen were carried out. The aim of the ongoing investigation was to determine and quantitatively describe the predominant processes of hydrogen embrittlement and their influence on the short fatigue crack morphology and crack growth rate. In addition, simulations were carried out on the short fatigue crack growth, in order to develop a detailed insight into the hydrogen embrittlement mechanisms relevant for cyclic loading conditions. It was found that a lower nickel content and a higher martensite content of the samples led to a higher susceptibility to hydrogen embrittlement. In addition, crack propagation and crack path could be simulated well with the simulation model.

Al and Si Influences on Hydrogen Embrittlement of Carbide-Free Bainitic Steel

Directory of Open Access Journals (Sweden)

Yanguo Li

2013-01-01

Full Text Available A first-principle method based on the density functional theory was applied to investigate the Al and Si influences on the hydrogen embrittlement of carbide-free bainitic steel. The hydrogen preference site, binding energy, diffusion behaviour, and electronic structure were calculated. The results showed that hydrogen preferred to be at the tetrahedral site. The binding energy of the cell with Si was the highest and it was decreased to be the worst by adding hydrogen. The diffusion barrier of hydrogen in the cell containing Al was the highest, so it was difficult for hydrogen to diffuse. Thus, hydrogen embrittlement can be reduced by Al rather than Si.

Helium cryogenics

CERN Document Server

Van Sciver, Steven W

2012-01-01

Twenty five years have elapsed since the original publication of Helium Cryogenics. During this time, a considerable amount of research and development involving helium fluids has been carried out culminating in several large-scale projects. Furthermore, the field has matured through these efforts so that there is now a broad engineering base to assist the development of future projects. Helium Cryogenics, 2nd edition brings these advances in helium cryogenics together in an updated form. As in the original edition, the author's approach is to survey the field of cryogenics with emphasis on helium fluids. This approach is more specialized and fundamental than that contained in other cryogenics books, which treat the associated range of cryogenic fluids. As a result, the level of treatment is more advanced and assumes a certain knowledge of fundamental engineering and physics principles, including some quantum mechanics. The goal throughout the work is to bridge the gap between the physics and engineering aspe...

Liquid helium

CERN Document Server

Atkins, K R

1959-01-01

Originally published in 1959 as part of the Cambridge Monographs on Physics series, this book addresses liquid helium from the dual perspectives of statistical mechanics and hydrodynamics. Atkins looks at both Helium Three and Helium Four, as well as the properties of a combination of the two isotopes. This book will be of value to anyone with an interest in the history of science and the study of one of the universe's most fundamental elements.

Transition temperature of embrittlement of steel 11 474.1 welded joint

International Nuclear Information System (INIS)

Petrikova, A.; Cocher, M.

1987-01-01

The results are presented of tests of notch toughness in dependence on temperature for steel 11 474.1 used for the manufacture of steam separators, in the area of a joint welded using an automatic submerged-arc welding machine with pre-heating at 200 to 250 degC. After welding, the welded joints were annealed for reduced stress for 160 minutes at a temperature of 600 to 650 degC and left to cool off in the furnace. The obtained results show that: (1) critical embrittlement temperature for the welded joint and the given welding technology ranges within -20 and -13 degC; (2) critical embrittlement temperature following heat ageing is shifted to positive temperature values; (3) pressure tests of the steam separator jacket made of steel 11 474.1 may in the process of production be carried out at a minimal wall temperature of 17 degC; (4) in case a pressure test has to be made after the equipment has been in operation for a certain period of time the test will probably have to be made at temperatures higher than 20 degC; (5) further tests will have to be made at temperatures higher than 20 degC in order to determine critical embrittlement temperatures after ageing. (J.B.). 7 figs., 2 tabs., 5 refs

Non-destructive evaluation of thermal aging embrittlement of duplex stainless steels

International Nuclear Information System (INIS)

Yi, Y.S.; Tomobe, T.; Watanabe, Y.; Shoji, T.

1993-01-01

The non-destructive evaluation procedure for detecting thermal aging embrittlement of cast duplex stainless steels has been investigated. As a novel measurement technique for the thermal aging embrittlement, an electrochemical method was used and anodic polarization behaviors were measured on new, service exposed, and laboratory aged materials and then were compared with the results of the mechanical tests and microstructural changes. During the polarization experiments performed in potassium hydroxide solution (KOH), M 23 C 6 carbides on phase boundary were preferentially dissolved, which was comfirmed by the SEM after polarization measurements. The preferential dissolution of M 23 C 6 carbides were obtained. Also, the non-destructive measurement and evaluation method of spinodal decomposition, which has been known as the primary mechanism of embrittlement inferrite phase, was reviewed. When the materials, where spinodal decomposition occurred, were polarized in an acetic acid solution (CH 3 COOH), larger critical anodic current densities were observed than those observed on new materials, and these results were consistent with the result of the microhardness measurement. Concerning these polarization results, a critical electric charge, which was required for stable passive films in passive metals, was defined and the relationship between the microstructural changes and this charge amount was reviewed under various polarization conditions in order to verify the polarization mechanism of the spinodally decomposed ferrite phase

Evaluation on thermal aging embrittlement of cast stainless steel components in domestic PWRs

Energy Technology Data Exchange (ETDEWEB)

Lee, Bong Sang; Hwa, Hong Jun; Chi, Se Hwan; Ryu, Woo Seog; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of thermal aging embrittlement of cast stainless steel components in PWRs. Cast stainless steel is being widely used in PWRs including primary piping. This material shows the reduction of fracture toughness during operating life due to high temperature. Micromechanisms and kinetics are summarized to improve the materials properties. The reduction of toughness due to thermal embrittlement in domestic reactors are predicted based on each chemical composition until the end of plant life time. Substantial degradation was predicted in some components during plant life time. (Author) 26 refs., 19 figs., 11 tabs.

Hydrogen embrittlement considerations in niobium-base alloys for application in the ITER divertor

International Nuclear Information System (INIS)

Peterson, D.T.; Hull, A.B.; Loomis, B.A.

1991-01-01

The ITER divertor will be subjected to hydrogen from aqueous corrosion by the coolant and by transfer from the plasma. Global hydrogen concentrations are one factor in assessing hydrogen embrittlement but local concentrations affected by source fluxes and thermotransport in thermal gradients are more important considerations. Global hydrogen concentrations is some corrosion- tested alloys will be presented and interpreted. The degradation of mechanical properties of Nb-base alloys due to hydrogen is a complex function of temperature, hydrogen concentration, stresses and alloy composition. The known tendencies for embrittlement and hydride formation in Nb alloys are reviewed

Study and prediction model on low temperature aging embrittlement in duplex stainless steels

International Nuclear Information System (INIS)

Sanchez, L.; Gutierrez-Solana, F.

1997-01-01

Within the framework of a general study on low temperature (280-400 degree centigree) aging embrittlement in duplex stainless steels, a relationship has been obtained between aging, measured from ferrite hardness evolution, and bulk materials embrittlement, determined from fracture toughness and fracture impact tests. The existing correlation between the increase in ferrite hardness and its percentage presence in the fracture path supports this relationship and results in the development of a prediction design model which provides the fracture resistance curves, for any aging level, based on the chemical composition and the steel's properties in an unaged state. (Author)

A wide-range embrittlement trend curve for western RPV steels

International Nuclear Information System (INIS)

Kirk, M.T.

2011-01-01

Embrittlement trend curves (ETCs) are used to estimate neutron irradiation embrittlement as a function of both exposure (fluence, flux, temperature, ...) and composition variables. ETCs provide information needed to assess the structural integrity of operating nuclear reactors, and to determine their suitability for continued safe operation. Past efforts on ETC development in the United States have used data drawn from domestic licensees. While this approach has addressed past needs well, future needs such as power up-rates, license extensions to 60 years and beyond, and the use of low copper materials in new reactors produce future operating conditions for the US reactor fleet that may differ from past experience, suggesting that data from sources other than licensee surveillance programs may be needed. In this paper we draw together embrittlement data expressed in terms of ΔT41J and ΔYS from a wide variety of data sources as a first step in examining future embrittlement trends. We develop a 'wide range' ETC based on a collection of over 2500 data. We assess how well this ETC models the whole database, as well as significant data subsets. Comparisons presented herein indicate that a single algebraic model, denoted WR-C(5), represents reasonably well both the trends evident in the data overall as well as trends exhibited by four special data subsets. The WR-C(5) model indicates the existence of trends in high fluence data (Φ > 2-3*10 19 n/cm 2 , E > 1 MeV) that are not as apparent in the US surveillance data due to the limited quantity of ΔT30 data measured at high fluence in this dataset. Additionally, WR-C(5) models well the trends in both test and power reactor data despite the fact it has not term to account for flux. It is suggested that one appropriate use of the WR-C(5) trend curve may include the design irradiation studies to validate or refute the findings presented herein. Additionally, WR-C(5) could be used, along with other information (e.g., other

CALCULATED REGENERATOR PERFORMANCE AT 4 K WITH HELIUM-4 AND HELIUM-3

International Nuclear Information System (INIS)

Radebaugh, Ray; Huang Yonghua; O'Gallagher, Agnes; Gary, John

2008-01-01

The helium-4 working fluid in regenerative cryocoolers operating with the cold end near 4 K deviates considerably from an ideal gas. As a result, losses in the regenerator, given by the time-averaged enthalpy flux, are increased and are strong functions of the operating pressure and temperature. Helium-3, with its lower boiling point, behaves somewhat closer to an ideal gas in this low temperature range and can reduce the losses in 4 K regenerators. An analytical model is used to find the fluid properties that strongly influence the regenerator losses as well as the gross refrigeration power. The thermodynamic and transport properties of helium-3 were incorporated into the latest NIST regenerator numerical model, known as REGEN3.3, which was used to model regenerator performance with either helium-4 or helium-3. With this model we show how the use of helium-3 in place of helium-4 can improve the performance of 4 K regenerative cryocoolers. The effects of operating pressure, warm-end temperature, and frequency on regenerators with helium-4 and helium-3 are investigated and compared. The results are used to find optimum operating conditions. The frequency range investigated varies from 1 Hz to 30 Hz, with particular emphasis on higher frequencies

Embrittlement of the alloy U 7.5 Nb 2.5 Zr by gaseous oxygen and hydrogen

International Nuclear Information System (INIS)

Lepoutre, D.; Nomine, A.M.; Miannay, D.

1981-04-01

Embrittlement of the alloy uranium 7.5 niobium 2.5 zirconium in gaseous oxygen and hydrogen versus stress intensity, temperature and pressure is studied using rupture mechanics. Cracking speed is determined. In oxygen, only cracks are produced and embrittlement is due to oxidation. In hydrogen at high pressure an hydride is formed and at low pressure cracks are produced but the mechanism is not identified [fr

High temperature service embrittlement of EUROFER´97 steel

Czech Academy of Sciences Publication Activity Database

Stratil, Luděk; Hadraba, Hynek; Dlouhý, Ivo

2010-01-01

Roč. 1, č. 2 (2010), s. 142-145 ISSN 1335-1532. [Fraktografia 2009. Stará Lesná, 08.11.2009-11.11.2009] R&D Projects: GA ČR GA106/08/1397; GA AV ČR 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : Eurofer´97 * isothermal ageing * embrittlement * impact properties Subject RIV: JL - Materials Fatigue, Friction Mechanics

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

Comparison of embrittlement trend curves to high fluence surveillance results

International Nuclear Information System (INIS)

Bogaert, A.S.; Gerard, R.; Chaouadi, R.

2011-01-01

In the regulatory justification of the integrity of the reactor pressure vessels (RPV) for long term operation, use is made of predictive formulas (also called trend curves) to evaluate the RPV embrittlement (expressed in terms of RTNDT shifts) in function of fluence, chemical composition and in some cases temperature, neutron flux or product form. It has been shown recently that some of the existing or proposed trend curves tend to underpredict high dose embrittlement. Due to the scarcity of representative surveillance data at high dose, some test reactor results were used in these evaluations and raise the issue of representativeness of the accelerated test reactor irradiations (dose rate effects). In Belgium the surveillance capsules withdrawal schedule was modified in the nineties in order to obtain results corresponding to 60 years of operation or more with the initial surveillance program. Some of these results are already available and offer a good opportunity to test the validity of the predictive formulas at high dose. In addition, advanced surveillance methods are used in Belgium like the Master Curve, increased tensile tests, and microstructural investigations. These techniques made it possible to show the conservatism of the regulatory approach and to demonstrate increased margins, especially for the first generation units. In this paper the surveillance results are compared to different predictive formulas, as well as to an engineering hardening model developed at SCK.CEN. Generally accepted property-to-property correlations are critically revisited. Conclusions are made on the reliability and applicability of the embrittlement trend curves. (authors)

The influence of second-phase dispersion on environmental embrittlement of Ni3(Si,Ti) alloys

International Nuclear Information System (INIS)

Takasugi, T.; Hanada, S.

1999-01-01

Some quaternary Ni 3 (Si,Ti) alloyed with transition elements V, Nb, Zr and Hf was prepared beyond their maximum solubility limits to investigate the effect of second-phase dispersion on moisture-induced embrittlement. V-added Ni 3 (Si,Ti) alloy contained ductile fcc-type Ni solid solution as the second-phase, while Nb-, Zr- and Hf-added Ni 3 (Si,Ti) alloys contained hard dispersion compounds as the second-phase. V- and Nb-added Ni 3 (Si,Ti) alloys did not display reduced tensile elongation in air, indicating that their second phases have the effect of suppressing the moisture-induced embrittlement. Possible mechanisms for the beneficial effect by the second phase on the moisture-induced embrittlement of V- and Nb-added Ni 3 (Si,Ti) alloys are discussed in association with hydrogen behavior and deformation property in the constituent phases or at matrix/second-phase interface

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

Quantitative evaluation of rejuvenators to restore embrittlement temperatures in oxidized asphalt mixtures using acoustic emission

Science.gov (United States)

Sun, Zhe; Farace, Nicholas; Arnold, Jacob; Behnia, Behzad; Buttlar, William G.; Reis, Henrique

2015-03-01

Towards developing a method capable to assess the efficiency of rejuvenators to restore embrittlement temperatures of oxidized asphalt binders towards their original, i.e., unaged values, three gyratory compacted specimens were manufactured with mixtures oven-aged for 36 hours at 135 °C. In addition, one gyratory compacted specimen manufactured using a short-term oven-aged mixture for two hours at 155 °C was used for control to simulate aging during plant production. Each of these four gyratory compacted specimens was then cut into two cylindrical specimen 5 cm thick for a total of six 36-hour oven-aged specimens and two short term aging specimens. Two specimens aged for 36 hours and the two short-term specimens were then tested using an acoustic emission approach to obtain base acoustic emission response of short-term and severely-aged specimens. The remaining four specimens oven-aged for 36 hours were then treated by spreading their top surface with rejuvenator in the amount of 10% of the binder by weight. These four specimens were then tested using the same acoustic emission approach after two, four, six, and eight weeks of dwell time. It was observed that the embrittlement temperatures of the short-term aged and severely oven-aged specimens were -25 °C and - 15 °C, respectively. It was also observed that after four weeks of dwell time, the rejuvenator-treated samples had recuperated the original embrittlement temperatures. In addition, it was also observed that the rejuvenator kept acting upon the binder after four weeks of dwell time; at eight weeks of dwell time, the specimens had an embrittlement temperature about one grade cooler than the embrittlement temperature corresponding to the short-term aged specimen.

New helium spectrum variable and a new helium-rich star

International Nuclear Information System (INIS)

Walborn, N.R.

1974-01-01

HD 184927, known previously as a helium-rich star, has been found to have a variable helium spectrum; the equivalent widths of five He I lines are larger by an average of 46 percent on a 1974 spectrogram than on one obtained with the same equipment in 1970. HD 186205 has been found to be a new, pronounced helium-rich star. (auth)

The effects of composition on the environmental embrittlement of Fe{sub 3}Al alloys

Energy Technology Data Exchange (ETDEWEB)

Alven, D.A.; Stoloff, N.S. [Rensselaer Polytechnic Inst., Troy, NY (United States)

1997-12-01

This paper reviews recent research on embrittlement of iron aluminides at room temperature brought about by exposure to moisture or hydrogen. The tensile and fatigue crack growth behavior of several Fe-28Al-5Cr alloys with small additions of Zr and C are described. It will be shown that fatigue crack growth behavior is dependent on composition, environment, humidity level, and frequency. Environments studied include vacuum, oxygen, hydrogen gas, and moist air. All cases of embrittlement are ultimately traceable to the interaction of hydrogen with the crack tip.

Neutron embrittlement of the Kozloduy NPP unit 1 reactor

International Nuclear Information System (INIS)

Vodenicharov, S.; Kamenova, Tz.

1996-01-01

Activities made in the period 1989-1996 according to the Program for metal state monitoring of the Kozloduy NPP Unit 1 are described. Data on P and Cu content in the welded joint 4 are reported. Determination is made by wet chemical analysis of shavings taken out from the inner side of the wall, direct spectral analysis of the vessel itself and spectroscopy of the inner and outer side of 6 templates. The results obtained from 4 different study teams showed a good agreement. The real average P content is 0.046% and tends to diminish in depth. Microstructural investigation does not show any expressed inter-crystalline mechanism of brittle failure at low temperatures. The data on real P and Cu content, as well as the experimental values of the initial critical temperature of embrittlement (Tk o ), the residual part of temperature shift (Tk r ) and the re-embrittlement temperature after annealing at 475 o (Tk) allow to predict the change in Tk o of the joint 4 during the next refueling cycles. The measured low value of Tk after 18-th refueling cycle is considerably lower than that forecasted by lateral re-embrittlement law. This means that the forecasting of Tk for the next cycles is made with big enough conservatisms, and that a second annealing of the vessel until 26-th cycle is not necessary. So according to the most conservative estimate, the Unit 1 can operate safely until the end of the 26-th refueling cycle. It is also concluded, that in terms of radiation degradation of the vessel metal the operation life time of the Unit 1 can reach and exceed the designed one. 2 tab., 7 ref

The modelling of irradiation embrittlement in submerged-arc welds

International Nuclear Information System (INIS)

Bolton, C.J.; Buswell, J.T.; Jones, R.B.; Moskovic, R.; Priest, R.H.

1996-01-01

Until very recently, the irradiation embrittlement behavior of submerged-arc welds has been interpreted in terms of two mechanisms, namely a matrix damage component and an additional component due to the irradiation-enhanced production of copper-rich precipitates. However, some of the weld specimens from a recent accelerated re-irradiation experiment have shown high Charpy shifts which exceeded the values expected from the measured shift in yield stress. Microstructural examination has revealed the occurrence of intergranular fracture (IGF) in these specimens, accompanied by grain boundary segregation of phosphorus. Theoretical models were developed to predict the parametric dependence of irradiation-enhanced phosphorus segregation on experimental variables. Using these parametric forms, along with the concept of a critical level of segregation for the onset of IGF instead of cleavage, a three mechanism trend curve has been developed. The form of this trend curve, taking into account IGF as well as matrix and copper embrittlement, is thus mechanistically based. The constants in the equation, however, are obtained by a statistical fit to the actual Charpy shift database

Void migration in fusion materials

International Nuclear Information System (INIS)

Cottrell, G.A.

2002-01-01

Neutron irradiation in a fusion power plant will cause helium bubbles and voids to form in the armour and blanket structural materials. If sufficiently large densities of such defects accumulate on the grain boundaries of the materials, the strength and the lifetimes of the metals will be reduced by helium embrittlement and grain boundary failure. This Letter discusses void migration in metals, both by random Brownian motion and by biassed flow in temperature gradients. In the assumed five-year blanket replacement time of a fusion power plant, approximate calculations show that the metals most resilient to failure are tungsten and molybdenum, and marginally vanadium. Helium embrittlement and grain boundary failure is expected to be more severe in steel and beryllium

Void migration in fusion materials

Science.gov (United States)

Cottrell, G. A.

2002-04-01

Neutron irradiation in a fusion power plant will cause helium bubbles and voids to form in the armour and blanket structural materials. If sufficiently large densities of such defects accumulate on the grain boundaries of the materials, the strength and the lifetimes of the metals will be reduced by helium embrittlement and grain boundary failure. This Letter discusses void migration in metals, both by random Brownian motion and by biassed flow in temperature gradients. In the assumed five-year blanket replacement time of a fusion power plant, approximate calculations show that the metals most resilient to failure are tungsten and molybdenum, and marginally vanadium. Helium embrittlement and grain boundary failure is expected to be more severe in steel and beryllium.

Helium the disappearing element

CERN Document Server

Sears, Wheeler M

2015-01-01

The subject of the book is helium, the element, and its use in myriad applications including MRI machines, particle accelerators, space telescopes, and of course balloons and blimps. It was at the birth of our Universe, or the Big Bang, where the majority of cosmic helium was created; and stellar helium production continues. Although helium is the second most abundant element in the Universe, it is actually quite rare here on Earth and only exists because of radioactive elements deep within the Earth. This book includes a detailed history of the discovery of helium, of the commercial industry built around it, how the helium we actually encounter is produced within the Earth, and the state of the helium industry today. The gas that most people associate with birthday party balloons is running out. “Who cares?” you might ask. Well, without helium, MRI machines could not function, rockets could not go into space, particle accelerators such as those used by CERN could not operate, fiber optic cables would not...

Significance of rate of work hardening in tempered martensite embrittlement

International Nuclear Information System (INIS)

Pietikainen, J.

1995-01-01

The main explanations for tempered martensite embrittlement are based on the effects of impurities and cementite precipitation on the prior austenite grain boundaries. There are some studies where the rate of work hardening is proposed as a potential reason for the brittleness. One steel was studied by means of a specially developed precision torsional testing device. The test steel had a high Si and Ni content so ε carbide and Fe 3 C appear in quite different tempering temperature ranges. The M S temperature is low enough so that self tempering does not occur. With the testing device it was possible to obtain the true stress - true strain curves to very high deformations. The minimum toughness was always associated with the minimum of rate of work hardening. The change of deformed steel volume before the loss of mechanical stability is proposed as at least one reason for tempered martensite embrittlement. The reasons for the minimum of the rate of work hardening are considered. (orig.)

Embrittlement of MISSE 5 Polymers After 13 Months of Space Exposure

Science.gov (United States)

Guo, Aobo; Yi, Grace T.; Ashmead, Claire C.; Mitchell, Gianna G.; deGroh, Kim K.

2012-01-01

Understanding space environment induced degradation of spacecraft materials is essential when designing durable and stable spacecraft components. As a result of space radiation, debris impacts, atomic oxygen interaction, and thermal cycling, the outer surfaces of space materials degrade when exposed to low Earth orbit (LEO). The objective of this study was to measure the embrittlement of 37 thin film polymers after LEO space exposure. The polymers were flown aboard the International Space Station and exposed to the LEO space environment as part of the Materials International Space Station Experiment 5 (MISSE 5). The samples were flown in a nadir-facing position for 13 months and were exposed to thermal cycling along with low doses of atomic oxygen, direct solar radiation and omnidirectional charged particle radiation. The samples were analyzed for space-induced embrittlement using a bend-test procedure in which the strain necessary to induce surface cracking was determined. Bend-testing was conducted using successively smaller mandrels to apply a surface strain to samples placed on a semi-suspended pliable platform. A pristine sample was also tested for each flight sample. Eighteen of the 37 flight samples experienced some degree of surface cracking during bend-testing, while none of the pristine samples experienced any degree of cracking. The results indicate that 49 percent of the MISSE 5 thin film polymers became embrittled in the space environment even though they were exposed to low doses (approx.2.75 krad (Si) dose through 127 mm Kapton) of ionizing radiation.

Liquid helium target

International Nuclear Information System (INIS)

Fujii, Y.; Kitami, T.; Torikoshi, M.

1984-12-01

A liquid helium target system has been built and used for the experiment on the reaction 4 He(γ, p). The target system has worked satisfactorily; the consumption rate of liquid helium is 360 ml/h and the cryogenic system retains liquid helium for about ten hours. The structure, operation and performance of the target system are reported. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Susceptibility of 2 1/4 Cr-1Mo steel to liquid metal induced embrittlement by lithium-lead solutions

International Nuclear Information System (INIS)

Eberhard, B.A.; Edwards, G.R.

1984-08-01

An investigation has been conducted on the liquid metal induced embrittlement susceptibility of 2 1/4Cr-1Mo steel exposed to lithium and 1a/o lead-lithium at temperatures between 190 0 C and 525 0 C. This research was part of an ongoing effort to evaluate the compatibility of liquid lithium solutions with potential fusion reactor containment materials. Of particular interest was the microstructure present in a weld heat-affected zone, a microstructure known to be highly susceptible to corrosive attack by liquid lead-lithium solutions. Embrittlement susceptibility was determined by conducting tension tests on 2 1/4Cr-1Mo steel exposed to an inert environment as well as to a lead-lithium liquid and observing the change in tensile behavior. The 2 1/4Cr-1Mo steel was also given a base plate heat treatment to observe its embrittlement susceptibility to 1a/o lead-lithium. The base plate microstructure was severely embrittled at temperatures less than 500 0 C. Tempering the base plate was effective in restoring adequate ductility to the steel

A liquid helium saver

International Nuclear Information System (INIS)

Avenel, O.; Der Nigohossian, G.; Roubeau, P.

1976-01-01

A cryostat equipped with a 'liquid helium saver' is described. A mass flow rate M of helium gas at high pressure is injected in a counter-flow heat exchanger extending from room to liquid helium temperature. After isenthalpic expansion through a calibrated flow impedance this helium gas returns via the low pressure side of the heat exchanger. The helium boil-off of the cryostat represents a mass flow rate m, which provides additional precooling of the incoming helium gas. Two operating regimes appear possible giving nearly the same efficiency: (1) high pressure (20 to 25 atm) and minimum flow (M . L/W approximately = 1.5) which would be used in an open circuit with helium taken from a high pressure cylinder; and (2) low pressure (approximately = 3 atm), high flow (M . L/W > 10) which would be used in a closed circuit with a rubber diaphragm pumping-compressing unit; both provide a minimum theoretical boil-off factor of about 8%. Experimental results are reported. (U.K.)

Design and use of the Embrittlement Data Base (EDB)

International Nuclear Information System (INIS)

Stallmann, F.W.

1987-01-01

The architecture of the Embrittlement Data Base (EDB) is described. This data base contains a comprehensive collection of experimental data related to irradiations of reactor pressure vessel steels in surveillance capsules and test reactors. Software is being developed for easy retrieval and analysis of the data. Data and software will be made available to interested parties on a cooperative basis

Canada's helium output rising fast

Energy Technology Data Exchange (ETDEWEB)

1966-12-01

About 12 months from now, International Helium Limited will be almost ready to start up Canada's second helium extraction plant at Mankota, in Saskatchewan's Wood Mountain area about 100 miles southwest of Moose Jaw. Another 80 miles north is Saskatchewan's (and Canada's) first helium plant, operated by Canadian Helium and sitting on a gas deposit at Wilhelm, 9 miles north of Swift Current. It contains almost 2% helium, some COD2U, and the rest nitrogen. One year in production was apparently enough to convince Canadian Helium that the export market (it sells most of its helium in W. Europe) can take a lot more than it's getting. Construction began this summer on an addition to the Swift Current plant that will raise its capacity from 12 to 36MMcf per yr when it goes on stream next spring. Six months later, International Helium's 40 MMcf per yr plant to be located about 4 miles from its 2 Wood Mountain wells will double Canada's helium output again.

Dependence of hydrogen-induced lattice defects and hydrogen embrittlement of cold-drawn pearlitic steels on hydrogen trap state, temperature, strain rate and hydrogen content

International Nuclear Information System (INIS)

Doshida, Tomoki; Takai, Kenichi

2014-01-01

The effects of the hydrogen state, temperature, strain rate and hydrogen content on hydrogen embrittlement susceptibility and hydrogen-induced lattice defects were evaluated for cold-drawn pearlitic steel that absorbed hydrogen in two trapping states. Firstly, tensile tests were carried out under various conditions to evaluate hydrogen embrittlement susceptibility. The results showed that peak 2 hydrogen, desorbed at temperatures above 200 °C as determined by thermal desorption analysis (TDA), had no significant effect on hydrogen embrittlement susceptibility. In contrast, hydrogen embrittlement susceptibility increased in the presence of peak 1 hydrogen, desorbed from room temperature to 200 °C as determined by TDA, at temperatures higher than −30 °C, at lower strain rates and with higher hydrogen content. Next, the same effects on hydrogen-induced lattice defects were also evaluated by TDA using hydrogen as a probe. Peak 2 hydrogen showed no significant effect on either hydrogen-induced lattice defects or hydrogen embrittlement susceptibility. It was found that hydrogen-induced lattice defects formed under the conditions where hydrogen embrittlement susceptibility increased. This relationship indicates that hydrogen embrittlement susceptibility was higher under the conditions where the formation of hydrogen-induced lattice defects tended to be enhanced. Since hydrogen-induced lattice defects formed by the interaction between hydrogen and strain were annihilated by annealing at a temperature of 200 °C, they were presumably vacancies or vacancy clusters. One of the common atomic-level changes that occur in cold-drawn pearlitic steel showing higher hydrogen embrittlement susceptibility is the formation of vacancies and vacancy clusters

SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Metal Irradiation Embrittlement, annealing and Re-Embrittlement. Second Progress Report

International Nuclear Information System (INIS)

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

1999-07-01

The report gives the actual status of the contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement. Results from the reference testing of unirradiated material as well as the results of the CHIVAS-7 experiment are discussed

Reactor pressure vessel embrittlement management through EPRI-Developed material property databases

International Nuclear Information System (INIS)

Rosinski, S.T.; Server, W.L.; Griesbach, T.J.

1997-01-01

Uncertainties and variability in U.S. reactor pressure vessel (RPV) material properties have caused the U.S. Nuclear Regulatory Commission (NRC) to request information from all nuclear utilities in order to assess the impact of these data scatter and uncertainties on compliance with existing regulatory criteria. Resolving the vessel material uncertainty issues requires compiling all available data into a single integrated database to develop a better understanding of irradiated material property behavior. EPRI has developed two comprehensive databases for utility implementation to compile and evaluate available material property and surveillance data. RPVDATA is a comprehensive reactor vessel materials database and data management program that combines data from many different sources into one common database. Searches of the data can be easily performed to identify plants with similar materials, sort through measured test results, compare the ''best-estimates'' for reported chemistries with licensing basis values, quantify variability in measured weld qualification and test data, identify relevant surveillance results for characterizing embrittlement trends, and resolve uncertainties in vessel material properties. PREP4 has been developed to assist utilities in evaluating existing unirradiated and irradiated data for plant surveillance materials; PREP4 evaluations can be used to assess the accuracy of new trend curve predictions. In addition, searches of the data can be easily performed to identify available Charpy shift and upper shelf data, review surveillance material chemistry and fabrication information, review general capsule irradiation information, and identify applicable source reference information. In support of utility evaluations to consider thermal annealing as a viable embrittlement management option, EPRI is also developing a database to evaluate material response to thermal annealing. Efforts are underway to develop an irradiation

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)

Hydrogen embrittlement of high strength steel electroplated with zincâ  cobalt allo

OpenAIRE

Hillier, Elizabeth M. K.; Robinson, M. J.

2004-01-01

Slow strain rate tests were performed on quenched and tempered AISI 4340 steel to measure the extent of hydrogen embrittlement caused by electroplating with zincâ  cobalt alloys. The effects of bath composition and pH were studied and compared with results for electrodeposited cadmium and zincâ  10%nickel. It was found that zincâ  1%cobalt alloy coatings caused serious hydrogen embrittlement (EI 0.63); almost as severe as that of cadmium (EI 0.78). Baking cadmium plate...

Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

Science.gov (United States)

Kennedy, Daniel; Jaworski, Michael

2014-10-01

Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

An holistic approach to the problem of reactor ageing. [Pressure vessel embrittlement

Energy Technology Data Exchange (ETDEWEB)

Phythian, W.; McElroy, R.; Druce, S.; Kovan, D. (AEA Reactor Services, Harwell (United Kingdom))

1992-12-01

Understanding the process of ageing in reactors is essential to extending their lives beyond original design. To present a sound case -particularly regarding the level of embrittlement in reactor vessels due to radiation damage - an integrated approach using advanced assessment tools is needed. The techniques developed for the purpose involve, on the microscopic level, advanced neutron dosimetry and high resolution measurement techniques (eg advanced electron beam techniques and small angle neutron scattering) with which an analysis can be done of the radiation damage and the microstructural state of the steel test procedures (tensile, fracture toughness and Charpy impact) on standard and sub-sized specimens, the extent of radiation degradation can be characterised. finally, it is possible to predict how the degradation will evolve using physically-based models of embrittlement. (Author).

Hydrogen Embrittlement Mechanism in Fatigue Behaviour of Austenitic and Martensitic Stainless Steels

Directory of Open Access Journals (Sweden)

Brück Sven

2018-01-01

Full Text Available In the present study, the influence of hydrogen on the fatigue behaviour of the high strength martensitic stainless steel X3CrNiMo13-4 and the metastable austenitic stainless steels X2Crni19-11 with various nickel contents was examined in the low and high cycle fatigue regime. The focus of the investigations was the changes in the mechanisms of short crack propagation. The aim of the ongoing investigation is to determine and quantitatively describe the predominant processes of hydrogen embrittlement and their influence on the short fatigue crack morphology and crack growth rate. In addition, simulations were carried out on the short fatigue crack growth, in order to develop a detailed insight into the hydrogen embrittlement mechanisms relevant for cyclic loading conditions.

Estimation of RPV material embrittlement for Ukrainian NPP based on surveillance test data

International Nuclear Information System (INIS)

Revka, V.; Chyrko, L.; Chaikovsky, Yu.; Gulchuk, Yu.

2012-01-01

The WWER-1000 RPV material embrittlement has been evaluated using the surveillance test data for the nuclear power plant which is under operation in Ukraine. The RPV materials after the neutron (E > 0,5 MeV) irradiation up to fluence of 22,9.10 22 m -2 have been studied. Fracture toughness tests were performed using pre-cracked Charpy specimens for the beltline materials (base and weld metal). The maximum shift of T 0 reference temperature is equal to 44 o C. A radiation embrittlement rate, A F , for the RPV materials was estimated using the standard and reconstituted specimens. A comparison of the A F values has shown a good agreement between the specimen sets before and after reconstitution both for base and weld metal. Furthermore it has been revealed there is no nickel effect for the studied materials. In spite of the high nickel content the radiation embrittlement rate for weld metal is not higher than for base metal with low nickel content. Fracture toughness analysis has shown the Master curve shape describes well a temperature dependence of K Jc values. However a higher scatter of K Jc values is observed in comparison to 95 % tolerance bounds. (author)

Effect of Microstructure and Alloy Chemistry on Hydrogen Embrittlement of Precipitation-Hardened Ni-Based Alloys

Science.gov (United States)

Obasi, G. C.; Zhang, Z.; Sampath, D.; Morana, Roberto; Akid, R.; Preuss, M.

2018-04-01

The sensitivity to hydrogen embrittlement (HE) has been studied in respect of precipitation size distributions in two nickel-based superalloys: Alloy 718 (UNS N07718) and Alloy 945X (UNS N09946). Quantitative microstructure analysis was carried out by the combination of scanning and transmission electron microscopy and energy dispersive x-ray spectroscopy (EDS). While Alloy 718 is mainly strengthened by γ″, and therefore readily forms intergranular δ phase, Alloy 945X has been designed to avoid δ formation by reducing Nb levels providing high strength through a combination of γ' and γ″. Slow strain rate tensile tests were carried out for different microstructural conditions in air and after cathodic hydrogen (H) charging. HE sensitivity was determined based on loss of elongation due to the H uptake in comparison to elongation to failure in air. Results showed that both alloys exhibited an elevated sensitivity to HE. Fracture surfaces of the H precharged material showed quasi-cleavage and transgranular cracks in the H-affected region, while ductile failure was observed toward the center of the sample. The crack origins observed on the H precharged samples exhibited quasi-cleavage with slip traces at high magnification. The sensitivity is slightly reduced for Alloy 718, by coarsening γ″ and reducing the overall strength of the alloy. However, on further coarsening of γ″, which promotes continuous decoration of grain boundaries with δ phase, the embrittlement index rose again indicating a change of hydrogen embrittlement mechanism from hydrogen-enhanced local plasticity (HELP) to hydrogen-enhanced decohesion embrittlement (HEDE). In contrast, Alloy 945X displayed a strong correlation between strength, based on precipitation size and embrittlement index, due to the absence of any significant formation of δ phase for the investigated microstructures. For the given test parameters, Alloy 945X did not display any reduced sensitivity to HE compared with

Development of small punch tests for ductile-brittle transition temperature measurement of temper embrittled Ni-Cr steels

International Nuclear Information System (INIS)

Baik, J.M.; Kameda, J.; Buck, O.

1983-01-01

Small punch tests were developed to determine the ductile-brittle transition temperature of nickel-chromium (Ni-Cr) steels having various degrees of temper embrittlement and various microstructures. It was found that the small punch test clearly shows the ductile-brittle transition behavior of the temper-embrittled steels. The measured values were compared with those obtained from Charpy impact and uniaxial tensile tests. The effects of punch tip shape, a notch, and the strain rate on the ductile-brittle transition behavior were examined. It was found that the combined use of a notch, high strain rates, and a small punch tip strongly affects the ductile-brittle transition behavior. Considerable variations in the data were observed when the small punch tests were performed on coarse-grained steels. Several factors controlling embrittlement measurements of steels are discussed in terms of brittle fracture mechanisms

Embrittlement phenomenon of Ag core MP35N cable as lead conductor in medical device.

Science.gov (United States)

Wang, Ling; Li, Bernie; Zhang, Haitao

2013-02-01

Ag core MP35N (Ag/MP35N) wire has been used in lead electric conductor wires in the medical device industry for many years. Recently it was noticed that the combination of silver and MP35N restricts its wire drawing process. The annealing temperature in Ag/MP35N has to be lower than the melting temperature of pure Ag (960 °C), which cannot fully anneal MP35N. The lower annealing temperature results in a highly cold worked MP35N, which significantly reduces Ag/MP35N ductility. The embrittlement phenomenon of Ag/MP35N cable was observed in tension and bending deformation. The effect of the embrittlement on the wire flex fatigue life was evaluated using a newly developed flex fatigue testing method. The Ag/MP35N cable fatigue results was analyzed with a Coffin-Manson approach and compared to the MP35N cable fatigue results. The root causes of the Ag/Mp35N embrittlement phenomenon are discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Radiation embrittlement behavior of fine-grained molybdenum alloy with 0.2 wt%TiC addition

Energy Technology Data Exchange (ETDEWEB)

Kitsunai, Y. [Tohoku University (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan)]. E-mail: kurishi@imr.tohoku.ac.jp; Kuwabara, T. [Tohoku University (Japan); Narui, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Hasegawa, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Takida, T. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan); Takebe, K. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan)

2005-11-15

In order to elucidate the effects of pre-irradiation microstructures and irradiation conditions on radiation embrittlement and radiation-induced ductilization (RIDU), fine-grained Mo-0.2 wt%TiC specimens with high and low reduction rates in plastic working, which are designated as MTC-02H and MTC-02L, respectively, were prepared by powder metallurgical methods. The specimens were neutron irradiated to 0.1-0.15 dpa with controlled 1-cycle and 4-cycle heating between 573 and 773 K, and 473 and 673 K, respectively, in JMTR. Vickers microhardness and three-point bending impact tests and TEM microstructural examinations were made. The degree of radiation embrittlement, assessed by DBTT shift due to irradiation, was strongly dependent on the reduction rate and cycle number. The 4-cycle irradiation suppressed the radiation embrittlement compared with the 1-cycle irradiation, and the suppression was much more significant in MTC-02L than in MTC-02H. The observed behavior is discussed in connection with RIDU and microstructural evolution caused by the 4-cycle irradiation.

Study on the hydrogen embrittlement and corrosion of stainless steels used as NI/MHX battery containers

Energy Technology Data Exchange (ETDEWEB)

Chuang, H.J.; Chan, S.L.I. [National Taiwan University, Taipei (China); Chen, S.Y. [Chung Shan Institute of Science and Technology, Lung-Tan (China)

1998-07-01

Stainless steels are used as the containers for Nickel-metal hydride (Ni/MHx) batteries. In this work stainless steel 304, 304L, 316, 316L, 17-4PH and 430 were selected to study their relative susceptibility to hydrogen embrittlement and alkaline corrosion under battery environments. Comparisons were made by immersion test under different hydrogen pressure over the electrolyte, U-bend tests and slow strain rate tensile test with cathodic H{sub 2} charging. The results showed that high strength 17-4PH suffered severe corrosion after long time immersion in the electrolyte solution and were sensitive to hydrogen embrittlement after hydrogen charging. Ferritic 430 performed better than 17-4PH during immersion test but lost its ductility after hydrogen charging. All the austenitic steels (304, 304L, 316, 316L) were found to be suitable as the materials for Ni/MHx battery container, and the present tests can not discriminate their relative resistance to the corrosion and hydrogen embrittlement in the electrolyte. 5 refs.

Reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

1992-07-01

Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

Screw compressor system for industrial-scale helium refrigerators or industrial ammonia screw compressors for helium refrigeration systems; Schraubenkompressor-System fuer Helium-Grosskaelteanlage oder Ammoniak-Schraubenverdichter aus Industrieanwendungen fuer Helium-Kaelteanlagen

Energy Technology Data Exchange (ETDEWEB)

Fredrich, O.; Mosemann, D.; Zaytsev, D. [GEA Grasso GmbH Refrigeration Technology, Berlin (Germany)

2007-07-01

Material characteristics, requirements and measured data of ammonia and helium compression are compared. The compressor lines for industrial ammonia and helium refrigerators are presented, and important characteristics of the compressors are explained. The test stand for performance measurements with helium and ammonia is described, and results are presented. In spite of the different characteristics of the fluids, the compressor-specific efficiencies (supply characteristic, quality characteristic) were found to be largely identical. The values calculated for helium on the basis of NH3 test runs were found to be realistic, which means that the decades of experience with ammonia in industrial applications can be applied to helium compression as well. The design of screw compressor aggregates (skids) in industrial refrigeration is discussed and illustrated by examples. (orig.)

A review of formulas for predicting irradiation embrittlement of reactors vessel materials

International Nuclear Information System (INIS)

Petrequin, P.

1995-01-01

Formulas developed in different countries for predicting irradiation embrittlement of reactors vessel materials are presented. Results of predictions were compared with different data sets, from surveillance programmes or studies in test reactors, with different residual elements contents. Figs

Fracture behaviour of the 14Cr ODS steel exposed to helium and liquid lead

Science.gov (United States)

Hojna, Anna; Di Gabriele, Fosca; Hadraba, Hynek; Husak, Roman; Kubena, Ivo; Rozumova, Lucia; Bublikova, Petra; Kalivodova, Jana; Matejicek, Jiri

2017-07-01

This work describes the fracture behaviour of the 14Cr ODS steel produced by mechanical alloying process, after high temperature exposures. Small specimens were exposed to helium gas in a furnace at 720 °C for 500 h. Another set of specimens was exposed to flowing liquid lead in the COLONRI II loop at 650 °C for 1000 h. All specimens were tested for the impact and tensile behaviour. The impact test results are compared to other sets of specimens in the as received state and after isothermal annealing at 650 °C for 1000 h. The impact curves of the exposed materials showed positive shifts on the transition temperature. While the upper shelf value did not change in the Pb exposed ODS steel, it significantly increased in the He exposed one. The differences are discussed in terms of surface and subsurface microscopy observation. The embrittlement can be explained as the effect of a slight change in the grain boundary and size distribution combined with the depletion of sub-surface region from alloying elements forming oxide scale on the surface.

Self-trapping of helium in metals

International Nuclear Information System (INIS)

Wilson, W.D.; Bisson, C.L.; Baskes, M.I.

1981-01-01

Atomistic calculations are presented which demonstrate that helium atoms in a metal lattice are able to cluster with each other, producing vacancies and nearby self-interstitial defects. Even a small number of helium atoms is found to be sufficient to create these large distortions. As few as five interstitial helium can spontaneously produce a lattice vacancy and nearby self-interstitial. An eight-helium-atom cluster gives rise to two such defects, and 16 helium atoms to more than five self-interstitial vacancy pairs. It was noted that the self-interstitials prefer to agglomerate on the same ''side'' of the helium cluster rather than to spread themselves out uniformly. The binding energy of each additional helium atom to these clusters increases with helium concentration and the trap is apparently unsaturable. A rate theory using these atomistic binding energies has been used to calculate the kinetics of helium-bubble nucleation and growth. The results are consistent with measurements of the properties of helium resulting from tritium decay

Recent evaluation of 'wet' thermal annealing to resolve reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Server, W.L.; Biemiller, E.C.

1993-01-01

Prior to the decision to close the Yankee Rowe plant in 1992, a great deal of effort was expended in trying to resolve the degree of neutron embrittlement that the reactor pressure vessel had experienced after 30 years of operation. One mitigative measure that was examined in detail was the possibility of performing a relatively low temperature thermal anneal (at approximately 650 deg. F) to partially restore the original design level of mechanical properties of the reactor pressure vessel beltline region which were lost due to the neutron radiation exposure. This low temperature anneal was to involve heating of the primary coolant water using pump heat in a similar manner as that used to anneal the Belgian BR-3 reactor pressure vessel in the early 1980s. This 'wet' anneal was successful in recovering mechanical properties for the BR-3 vessel, but the extent of the recovery, as well as the rate of re-embrittlement after the anneal, were issues that were difficult to quantify since the exact reactor pressure vessel steels were not available for experimental verification. For the case of Yankee Rowe, material was available from past surveillance programs for at least one of the materials in the vessel, as well as materials obtained from various sources which could act as bounding surrogates. An irradiation /annealing/reirradiation program was developed to better quantify the degree of recovery and re-embrittlement for these materials, but this program was halted before significant test results were obtained. Prior to the initiation of the testing program, a review of past annealing data was performed and the data were scrutinized for direct relevance to the annealing response of the Yankee Rowe vessel. This paper discusses the results derived from this review. The results from the critical review of the past annealing data indicated that a 'wet' anneal of the Yankee Rowe vessel may have been successful in reducing the degree of embrittlement to the point that the

Evaluation of the current status of hydrogen embrittlement and stress-corrosion cracking in steels

Energy Technology Data Exchange (ETDEWEB)

Moody, N.R.

1981-12-01

A review of recent studies on hydrogen embrittlement and stress-corrosion cracking in steels shows there are several critical areas where data is either ambiguous, contradictory, or non-existent. A relationship exists between impurity segregation and hydrogen embrittlement effects but it is not known if the impurities sensitize a preferred crack path for hydrogen-induced failure or if impurity and hydrogen effects are additive. Furthermore, grain boundary impurities may enhance susceptibility through interactions with some environments. Some studies show that an increase in grain size increases susceptibility; at least one study shows an opposite effect. Recent work also shows that fracture initiates at different locations for external and internal hydrogen environments. How this influences susceptibility is unknown.

Convective mixing in helium white dwarfs

International Nuclear Information System (INIS)

Vauclair, G.; Fontaine, G.

1979-01-01

The conditions under which convective mixing episodes take place between the helium envelopes and the underlying carbon layers in helium-rich white dwarfs are investigated. It is found that, for essentially any value of the initial helium content less than the maximum mass a helium convection zone can have, mixing does occur, and leads, in the vast majority of cases, to an almost pure carbon superficial composition. Mixing products that show only traces of carbon while retaining helium-dominated envelopes are possible only if the initial helium content is quite close to the maximum possible mass of the helium convection zone. In the presence of turbulence, this restriction could be relaxed, however, and the helium-rich lambda4670 stars may possibly be explained in this fashion

Hydrogen embrittlement of the 22 Cr5 Ni austeno-ferritic stainless steel. Role of the microstructure

International Nuclear Information System (INIS)

Iacoviello, Francesco

1997-01-01

Austenitic-ferritic stainless steels are characterised by very good mechanical properties and by a high corrosion resistance, especially to stress-corrosion and to pitting. However, their duplex structure shows a sensitivity to hydrogen embrittlement. Among duplex stainless steels, the 22 Cr 5 Ni grade has gradually became the most used. In this work the tensile properties and the resistance to fatigue crack propagation of 22 Cr5 Ni duplex stainless steel have been analysed, with and without hydrogen charging, after it had been treated at temperatures ranging between 200-1050 deg. C for varying times. The heat treatment times and temperatures were chosen to characterise completely the effects of the different intermetallic and the carbide and nitride phases and to compare these results with those from the tensile tests and those in the literature. A technique for obtaining the hydrogen diffusion coefficient in the steel was optimised and was shown to be alternative to the permeation technique. Thermal analysis was used to determine the activation energy of the hydrogen traps in the steel. From the results the following conclusions were established: - Grain boundaries and dislocations have very little influence on the process of hydrogen diffusion. - The quantity of hydrogen absorbed depends in that any type of precipitate decrease the absorption. This decrease was probably due to changes in the diffusivity and solubility of hydrogen caused by the precipitation. - The charging with hydrogen caused a large decrease in ε m pc for the steel for all heat treatments temperature, except 1050 deg. C. At this temperature the effect was much less as the dislocation density was very low and the precipitates were now in solution. - Hydrogen charging of the steel did not affect the YS and the decrease in UTS produced depended on the microstructure. Use of the embrittlement index 'F' showed that spinodal decomposition and precipitation of G phase decrease hydrogen embrittlement

Regulatory aspects of radiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Randall, P.N.

1979-01-01

One purpose of this conference, is to re-examine the conventional wisdom about neutron radiation embrittlement and the methods used to counteract embrittlement in reactor vessels. Perhaps, there have been sufficient advances in fracture mechanics, core physics, dosimetry, and physical metallurgy to permit a forward step in the quantitative treatment of the subject. Certainly this would be consistent with the position of the U.S. Nuclear Regulatory Commission (the NRC) in general. ''There has been a continued evolution toward increased specificity.'' This statement appeared in the response prepared by the staff to a request from the Commission to explain how the staff decides to apply a new requirement and to whom, i.e., to back-fit or forward-fit-only or whatever. Pressure for increased specificity, i.e., for fleshing out general design criteria, comes from ''technical surprises'' in the form of operating experiences or from research information, and from attempts to improve our confidence in the safety of plants, especially new plants. Our goal is to have anticipated and evaluated all possible modes of failure with sufficient quantitativeness that the probability of failure can be estimated with some accuracy. Failing this, regulators demand large margins of safety to cover our ignorance

Friendly fermions of helium-three

International Nuclear Information System (INIS)

Leggatt, T.

1976-01-01

The importance of helium in showing up the effects of atomic indistinguishability and as a material by which to test some of the most fundamental principles of quantum mechanics is discussed. Helium not only remains liquid down to zero temperature but of the two isotopes helium-three has intrinsic spin 1/2 and should therefore obey the Pauli principle, while helium-four has spin zero and is expected to undergo Bose condensation. Helium-three becomes superfluid at temperatures of a few thousandths of a degree above absolute zero by the bulk liquid collecting its atoms into spinning pairs. There are three different superfluid phases, now conveniently called A, B and A 1 and each is characterised by a different behaviour of the spin and/or relative angular motion of the atoms composing the Cooper pairs. Problems surrounding the complicated physical system of helium-three are discussed. It is suggested that the combined coherence and directionality of superfluid helium-three should create some fascinating physics. (U.K.)

Fractography of hydrogen-embrittled iron-chromium-nickel alloys

International Nuclear Information System (INIS)

Caskey, G.R. Jr.

1980-01-01

Tensile specimens of iron-chromium-nickel base alloys were broken in either a hydrogen environment or in air following thermal charging with hydrogen. Fracture surfaces were examined by scanning electron microscopy. Fracture morphology of hydrogen-embrittled specimens was characterized by: changed dimple size, twin-boundary parting, transgranular cleavage, and intergranular separation. The nature and extent of the fracture mode changes induced by hydrogen varied systematically with alloy composition and test temperature. Initial microstructure developed during deformation processing and heat treating had a secondary influence on fracture mode

Helium turbo-expander with an alternator

International Nuclear Information System (INIS)

Akiyama, Yoshitane

1980-01-01

Study was made on a helium turbo-expander, the heart of helium refrigerator systems, in order to develop a system which satisfies the required conditions. A helium turbo-expander with externally pressurized helium gas bearings at the temperature of liquid nitrogen and an alternator as a brake have been employed. The essential difference between a helium turbo-expander and a nitrogen turbo-expander was clarified. The gas bearing lubricated with nitrogen at room temperature and the gas bearing lubricated with helium at low temperature were tested. The flow rate of helium in a helium refrigerator for a large superconducting magnet is comparatively small, therefore a helium turbine must be small, but the standard for large turbine design can be applied to such small turbine. Using the alternator as a brake, the turbo-expander was easily controllable electrically. The prototype turbo-expander was made, and the liquefaction test with it and MHD power generation test were carried out. (Kako, I.)

Effect of helium on the tensile properties of several vanadium alloys

International Nuclear Information System (INIS)

Braski, D.N.

1985-01-01

Specimens of three vanadium alloys were implanted with different levels of 3 He using the tritium trick and subsequently tensile tested at elevated temperatures. The V-15Cr-5Ti and V-3Ti-1Si specimens were embrittled by 3 He at a level of 150 at. ppm while the VANSTAR-7 specimens were not. The embrittlement appeared to be caused by extensive 3 He bubble networks and possibly precipitate particles on the grain boundaries; hopefully the bubble distribution can be changed (for better simulation) by altering parameters of the tritium trick. The results of the investigation also show that the embrittlement resistance of vanadium alloys can be improved by adjustment of their composition and/or microstructure. 1 reference, 4 figures, 1 table

Dissolution of alpha-prime precipitates in thermally embrittled S2205-duplex steels during reversion-heat treatment

Directory of Open Access Journals (Sweden)

V. Shamanth

2015-01-01

Full Text Available Duplex stainless steels offer an attractive combination of strength, corrosion resistance and cost. In annealed condition duplex steels will be in thermodynamically metastable condition but when they are subjected to intermediate homologous temperature of ∼475 °C and below significant embrittlement occurs, which is one of the key material degradation properties that limits its upper service temperature in many applications. Hence the present study is aimed to study the effect of reversion heat treatment and its time on mechanical properties of the thermally embrittled steel. The results showed that 60 min reversion heat treated samples were able to recover the mechanical properties which were very close to annealed properties because when the embrittled samples were reversion heat treated at an elevated temperature of 550 °C which is above the (α + α′ miscibility gap, the ferritic phase was homogenized again. In other words, Fe-rich α and Cr-rich α′ prime precipitates which were formed during ageing become thermodynamically unstable and dissolve inside the ferritic phase.

Accelerated aging embrittlement of cast duplex stainless steel: Activation energy for extrapolation

International Nuclear Information System (INIS)

Chung, H.M.; Chopra, O.K.

1989-05-01

Cast duplex stainless steels, used extensively in LWR systems for primary pressure boundary components such as primary coolant pipes, valves, and pumps, are susceptible to thermal aging embrittlement at reactor operating or higher temperatures. Since a realistic aging embrittlement for end-of-life or life-extension conditions (i.e., 32--50 yr of aging at 280--320 degree C) cannot be produced, it is customary to simulate the metallurgical structure by accelerated aging at ∼400 degree C. Over the past several years, extensive data on accelerated aging have been reported from a number of laboratories. The most important information from these studies is the activation energy, namely, the temperature dependence of the aging kinetics between 280 and 400 degree C, which is used to extrapolate the aging characteristics to reactor operating conditions. The activation energies (in the range of 18--50 kcal/mole) are, in general, sensitive to material grade, chemical composition, and fabrication process, and a few empirical correlations, obtained as a function of bulk chemical composition, have been reported. In this paper, a mechanistic understanding of the activation energy is described on the basis of the results of microstructural characterization of various heats of CF-3, -8, and -8M grades that were used in aging studies at different laboratories. The primary mechanism of aging embrittlement at temperatures between 280 and 400 degree C is the spinodal decomposition of the ferrite phase, and M 23 C 6 carbide precipitation on the ferrite/austenite boundaries is the secondary mechanism for high-carbon CF-8 grade. 20 refs., 10 figs., 3 tabs

Operating Manual of Helium Refrigerator (Rev. 2)

Energy Technology Data Exchange (ETDEWEB)

Song, K.M.; Son, S.H.; Kim, K.S.; Lee, S.K.; Kim, M.S. [Korea Electric Power Research Institute, Taejon (Korea)

2002-07-01

A helium refrigerator was installed as a supplier of 20K cold helium to the cryogenic distillation system of WTRF pilot plant. The operating procedures of the helium refrigerator, helium compressor and auxiliary apparatus are described for the safety and efficient operation in this manual. The function of the helium refrigerator is to remove the impurities from the compressed helium of about 250psig, to cool down the helium from ambient temperature to 20K through the heat exchanger and expansion engine and to transfer the cold helium to the cryogenic distillation system. For the smoothly operation of helium refrigerator, the preparation, the start-up, the cool-down and the shut-down of the helium refrigerator are described in this operating manual. (author). 3 refs., 14 tabs.

Hydrogen embrittlement and stress corrosion cracking in metals

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Cheong, Yong Mu; Im, Kyung Soo

2004-10-15

The objective of this report is to elucidate the mechanism for hydrogen embrittlement (HE) and stress corrosion cracking (SCC) in metals. To this end, we investigate the common features between delayed hydride cracking (DHC) in zirconium alloys and HE in metals with no precipitation of hydrides including Fe base alloys, Nickel base alloys, Cu alloys and Al alloys. Surprisingly, as with the crack growth pattern for the DHC in zirconium alloy, the metals mentioned above show a discontinuous crack growth, striation lines and a strong dependence of yield strength when exposed to hydrogen internally and externally. This study, for the first time, analyzes the driving force for the HE in metals in viewpoints of Kim's DHC model that a driving force for the DHC in zirconium alloys is a supersaturated hydrogen concentration coming from a hysteresis of the terminal solid solubility of hydrogen, not by the stress gradient, As with the crack growing only along the hydride habit plane during the DHC in zirconium alloys, the metals exposed to hydrogen seem to have the crack growing by invoking the dislocation slip along the preferential planes as a result of some interactions of the dislocations with hydrogen. Therefore, it seems that the hydrogen plays a role in inducing the slip only on the preferential planes so as to cause a strain localization at the crack tip. Sulfur in metals is detrimental in causing a intergranular cracking due to a segregation of the hydrogens at the grain boundaries. In contrast, boron in excess of 500 ppm added to the Ni3Al intermetallic compound is found to be beneficial in suppressing the HE even though further details of the mechanism for the roles of boron and sulfur are required. Carbon, carbides precipitating semi-continuously along the grain boundaries and the CSL (coherent site lattice) boundaries is found to suppress the intergranular stress corrosion cracking (IGSCC) in Alloy 600. The higher the volume fraction of twin boundaries, the

Hydrogen embrittlement and stress corrosion cracking in metals

International Nuclear Information System (INIS)

Kim, Young Suk; Cheong, Yong Mu; Im, Kyung Soo

2004-10-01

The objective of this report is to elucidate the mechanism for hydrogen embrittlement (HE) and stress corrosion cracking (SCC) in metals. To this end, we investigate the common features between delayed hydride cracking (DHC) in zirconium alloys and HE in metals with no precipitation of hydrides including Fe base alloys, Nickel base alloys, Cu alloys and Al alloys. Surprisingly, as with the crack growth pattern for the DHC in zirconium alloy, the metals mentioned above show a discontinuous crack growth, striation lines and a strong dependence of yield strength when exposed to hydrogen internally and externally. This study, for the first time, analyzes the driving force for the HE in metals in viewpoints of Kim's DHC model that a driving force for the DHC in zirconium alloys is a supersaturated hydrogen concentration coming from a hysteresis of the terminal solid solubility of hydrogen, not by the stress gradient, As with the crack growing only along the hydride habit plane during the DHC in zirconium alloys, the metals exposed to hydrogen seem to have the crack growing by invoking the dislocation slip along the preferential planes as a result of some interactions of the dislocations with hydrogen. Therefore, it seems that the hydrogen plays a role in inducing the slip only on the preferential planes so as to cause a strain localization at the crack tip. Sulfur in metals is detrimental in causing a intergranular cracking due to a segregation of the hydrogens at the grain boundaries. In contrast, boron in excess of 500 ppm added to the Ni3Al intermetallic compound is found to be beneficial in suppressing the HE even though further details of the mechanism for the roles of boron and sulfur are required. Carbon, carbides precipitating semi-continuously along the grain boundaries and the CSL (coherent site lattice) boundaries is found to suppress the intergranular stress corrosion cracking (IGSCC) in Alloy 600. The higher the volume fraction of twin boundaries, the more

Origin of intergranular embrittlement of Al alloys induced by Na and Ca segregation: Grain boundary weakening

International Nuclear Information System (INIS)

Lu Guanghong; Zhang Ying; Deng Shenghua; Wang Tianmin; Kohyama, Masanori; Yamamoto, Ryoichi; Liu Feng; Horikawa, Keitaro; Kanno, Motohiro

2006-01-01

Using a first-principles computational tensile test, we show that the ideal tensile strength of an Al grain boundary (GB) is reduced with both Na and Ca GB segregation. We demonstrate that the fracture occurs in the GB interface, dominated by the break of the interfacial bonds. Experimentally, we further show that the presence of Na or Ca impurity, which causes intergranular fracture, reduces the ultimate tensile strength when embrittlement occurs. These results suggest that the Na/Ca-induced intergranular embrittlement of an Al alloy originates mainly from the GB weakening due to the Na/Ca segregation

Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

Energy Technology Data Exchange (ETDEWEB)

Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

2011-10-01

Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 deg. C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

Relation between the conditions of helium ion implantation and helium void equilibrium parameters

International Nuclear Information System (INIS)

Neklyudov, I.M.; Rybalko, V.F.; Ruzhitskij, V.V.; Tolstolutskaya, G.D.

1981-01-01

The conditions of helium thermodynamic equilibrium in a system of voids produced by helium ion bombardment of a metal sample are studied. As an initial equation for description of the equilibrium the Clapeyron equation was used. The equation is obtained relating basic parameters of helium voids (average diameter and density) to irradiation parameters (dose, ion energy (straggling)) and properties of the metal (surface tension coefficient, yield strength). Comparison of the calculations with experimental data on helium in nickel found in literature shows that the equation yields satisfactory resutls for the dose range 1.10 16 -1x10 17 cm -2 and temperatures T [ru

Helium behaviour in nuclear glasses

International Nuclear Information System (INIS)

Fares, T.

2011-01-01

The present thesis focuses on the study of helium behavior in R7T7 nuclear waste glass. Helium is generated by the minor actinides alpha decays incorporated in the glass matrix. Therefore, four types of materials were used in this work. These are non radioactive R7T7 glasses saturated with helium under pressure, glasses implanted with 3 He + ions, glasses doped with curium and glasses irradiated in nuclear reactor. The study of helium solubility in saturated R7T7 glass has shown that helium atoms are inserted in the glass free volume. The results yielded a solubility of about 10 16 at. cm -3 atm. -1 . The incorporation limit of helium in this type of glass has been determined; its value amounted to about 2*10 21 at. cm -3 , corresponding to 2.5 at.%. Diffusion studies have shown that the helium migration is controlled by the single population dissolved in the glass free volume. An ideal diffusion model was used to simulate the helium release data which allowed to determine diffusion coefficients obeying to the following Arrhenius law: D = D 0 exp(-E a /kBT), where D 0 = 2.2*10 -2 and 5.4*10 -3 cm 2 s -1 and E a = 0.61 eV for the helium saturated and the curium doped glass respectively. These results reflect a thermally activated diffusion mechanism which seems to be not influenced by the glass radiation damage and helium concentrations studied in the present work (up to 8*10 19 at. g -1 , corresponding to 0.1 at.%). Characterizations of the macroscopic, structural and microstructural properties of glasses irradiated in nuclear reactor did not reveal any impact associated with the presence of helium at high concentrations. The observed modifications i.e. a swelling of 0.7 %, a decrease in hardness by 38 %, an increase between 8 and 34 % of the fracture toughness and a stabilization of the glass structure under irradiation, were attributed to the glass nuclear damage induced by the irradiation in reactor. Characterizations by SEM and TEM of R7T7 glasses implanted

Hydrogen embrittlement of metals. A bibliography with abstracts. Search period covered: 1964--August 1975

International Nuclear Information System (INIS)

Smith, M.F.

1975-10-01

The research covers the hydrogen embrittlement of both ferrous and nonferrous metals and alloys and includes nuclear technology, aircraft metallurgy, mechanical properties, testing, electroplating, fatigue, corrosion and fracture. Contains 230 abstracts

Radiation hardening and embrittlement of some refractory metals and alloys

International Nuclear Information System (INIS)

Fabritsiev, S.; Pokrovskyb

2007-01-01

Tungsten is proposed for application in the ITER divertor and limiter as plasma facing material. The tungsten operation temperature in the ITER divertor is relatively high. Hence, the ductile properties of tungsten will be controlled by the low temperature radiation embrittlement. The mechanism of radiation hardening and embrittlement under neutron irradiation at low temperature is well studied for FCC metals, in particular for copper. At the same time, low-temperature radiation hardening of BCC materials, in particular for refractory metals, is less studied. This study presents the results of investigation into radiation hardening and embrittlement of pure metals: W, Mo and Nb, and W-Re and Ta-4W alloys. The materials were in the annealed conditions. The specimens were irradiated in the SM-2 reactor to doses of 10 -4 -10 -1 dpa at 80 C and then tested for tension at 80 C. The study of the stress-strain curves of unirradiated specimens revealed a yield drop for W, Mo, Nb, Ta-4W, W-Re. After the yield drop some metals (Mo,Nb) retain their capability for strain hardening and demonstrate a high elongation (20-50%). Radiation hardening is maximum in Mo (∝400MPa) and minimum in Nb (∝100 MPa). In this case the dependence slope for Nb is similar to that for pure copper irradiated in SM-2 under the same conditions. Ii and Ta-4W have a higher slope. Measurement of electrical resistivity of irradiated specimens showed that for all materials it is increased monotonously with an increase in the irradiation dose. A minimum gain in electrical resistivity with a dose was observed for Nb (∝3% at 0.1 dpa). As for Mo it was essentially higher, i.e. ∝ 30%. The gain was maximum for W-Re alloy. Comparison of radiation hardening dose dependencies obtained in this study with the data for FCC metals (Cu) showed that in spite of the quantitative difference the qualitative behavior of these two classes of metals is similar. (orig.)

Cooling with Superfluid Helium

Energy Technology Data Exchange (ETDEWEB)

Lebrun, P; Tavian, L [European Organization for Nuclear Research, Geneva (Switzerland)

2014-07-01

The technical properties of helium II (‘superfluid’ helium) are presented in view of its applications to the cooling of superconducting devices, particularly in particle accelerators. Cooling schemes are discussed in terms of heat transfer performance and limitations. Large-capacity refrigeration techniques below 2 K are reviewed, with regard to thermodynamic cycles as well as process machinery. Examples drawn from existing or planned projects illustrate the presentation. Keywords: superfluid helium, cryogenics.

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

International Nuclear Information System (INIS)

Odette, G.R.; Lucas, G.E.; Wirth, B.; Liu, C.L.

1997-01-01

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

TAREG 2.01/00 Project, ''Validation of neutron embrittlement for VVER 1000 and 440/213 RPVs, with emphasis on integrity assessment''

International Nuclear Information System (INIS)

Ahlstrand, R.; Margolin, B.; Kostylev, V.; Yurchenko, E.; Akbashev, I.; Piminov, V.; Nikolaev, Y.; Koshkin, V.; Kharshenko, V.; Chyrko, L.; Bukhanov, V.; Comsa, O.

2012-01-01

The irradiation embrittlement and integrity of the VVER reactors has been an important issue in many EC supported TACIS and PHARE projects since 1990. In the EC annual program 2000 two TACIS projects (TAREG 2.01/00 and 2.01/03) were approved on the issue in order to improve the neutron irradiation embrittlement databases, elaborate new trend curves for the embrittlement and to assess the integrity of the RPVs (Reactor Pressure Vessel) by analysing PTS transients (Pressurized Thermal Shock) for some selected Russian and Ukrainian VVER 1000 and 440/213 NPPs. In this paper the TAREG 2.01/00 project is briefly described with some details from the twin project 2.01/03, which served as a materials testing project, providing inputs for the 1st project. As a result of the project new trend curves for neutron irradiation embrittlement were elaborated, based on upgraded and more reliable surveillance results databases. The PTS study shows that the integrity of the selected VVER RPVs can be ensured to the end of RPV design life. (author)

Bed system performance in helium circulation mode

Energy Technology Data Exchange (ETDEWEB)

Kim, Yean Jin; Jung, Kwang Jin; Ahn, Do Hee; Chung, Hong Suk [UST, Daejeon (Korea, Republic of); Kang, Hee Suk [KAERI, Daejeon (Korea, Republic of); Yun, Sei Hun [NFRI, Deajeon (Korea, Republic of)

2016-05-15

As a part of the International Thermonuclear Experimental Reactor (ITER) Project, We have conducted an experiment for storing hydrogen to depleted uranium and zirconium cobalt. The helium blanket effect has been observed in experiments using metal hydrides. The collapse of the hydrogen isotopes are accompanied by the decay heat and helium-3. Helium-3 dramatically reduces the hydrogen isotope storage capacity by surrounding the metal. This phenomenon is called a helium blanket effect. In addition the authors are working on the recovery and removal techniques of helium-3. In this paper, we discuss the equipment used to test the helium blanket effect and the results of a helium circulation experiment. The helium-3 produced surrounds the storage material surface and thus disturbs the reaction of the storage material and the hydrogen isotope. Even if the amount of helium-3 is small, the storage capacity of the SDS bed significantly drops. This phenomenon is the helium blanket effect. To resolve this phenomenon, a circulating loop was introduced. Using a circulating system, helium can be separated from the storage material. We made a helium loop that includes a ZrCo bed. Then using a metal bellows pump, we tested the helium circulation.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Multiscale Modeling of Grain Boundary Segregation and Embrittlement in Tungsten for Mechanistic Design of Alloys for Coal Fired Plants

Energy Technology Data Exchange (ETDEWEB)

Luo, Jian; Tomar, Vikas; Zhou, Naixie; Lee, Hongsuk

2013-06-30

Based on a recent discovery of premelting-like grain boundary segregation in refractory metals occurring at high temperatures and/or high alloying levels, this project investigated grain boundary segregation and embrittlement in tungsten (W) based alloys. Specifically, new interfacial thermodynamic models have been developed and quantified to predict high-temperature grain boundary segregation in the W-Ni binary alloy and W-Ni-Fe, W-Ni-Ti, W-Ni-Co, W-Ni-Cr, W-Ni-Zr and W-Ni-Nb ternary alloys. The thermodynamic modeling results have been experimentally validated for selected systems. Furthermore, multiscale modeling has been conducted at continuum, atomistic and quantum-mechanical levels to link grain boundary segregation with embrittlement. In summary, this 3-year project has successfully developed a theoretical framework in combination with a multiscale modeling strategy for predicting grain boundary segregation and embrittlement in W based alloys.

Helium supply demand in future years

International Nuclear Information System (INIS)

Laverick, C.

1975-01-01

Adequate helium will be available to the year 2000 AD to meet anticipated helium demands for present day applications and the development of new superconducting technologies of potential importance to the nation. It is almost certain that there will not be enough helium at acceptable financial and energy cost after the turn of the century to meet the needs of the many promising helium based technologies now under development. Serious consideration should be given to establishing priorities in development and application based upon their relative value to the country. In the first half of the next century, three ways of estimating helium demand lead to cumulative ranges of from 75 to 125 Gcf (economic study), 89 to 470 Gcf (projected national energy growth rates) and 154 to 328 Gcf (needs for new technologies). These needs contrast with estimated helium resources in natural gas after 2000 AD which may be as low as 10 or 126 Gcf depending upon how the federal helium program is managed and the nation's natural gas resources are utilized. The technological and financial return on a modest national investment in further helium storage and a rational long term helium program promises to be considerable

Surface electrons of helium films

International Nuclear Information System (INIS)

Studart, N.; Hipolito, O.

1986-01-01

Theoretical calculations of some properties of two-dimensional electrons on a liquid helium film adsorbed on a solid substrate are reviewed. We describe the spectrum of electron bound states on bulk helium as well on helium films. The correlational properties, such as the structure factor and correlation energy, are determined as functions of the film thickness for different types of substrates in the framework of a Generalized Random-Phase Approximation. The collective excitations of this system are also described. The results for electrons on the surface of thin films and bulk helium are easily obtained. we examine the electron interaction with the excitations of the liquid helium surface resulting in a new polaron state, which was observed very recently. The ground state energy and the effective mass of this polaron are determined by using the path-integral formalism and unitary-transformation method. Recent speculations about the phase diagram of electrons on the helium film are also discussed. (Author) [pt

Helium dilution refrigerator

International Nuclear Information System (INIS)

1973-01-01

A new system of continuous heat exchange for a helium dilution refrigerator is proposed. The 3 He effluent tube is concurrent with the affluent mixed helium tube in a vertical downward direction. Heat exchange efficiency is enhanced by placing in series a number of elements with an enlarged surface area

The Role of Hydrogen-Enhanced Strain-Induced Lattice Defects on Hydrogen Embrittlement Susceptibility of X80 Pipeline Steel

Science.gov (United States)

Hattori, M.; Suzuki, H.; Seko, Y.; Takai, K.

2017-08-01

Studies to date have not completely determined the factors influencing hydrogen embrittlement of ferrite/bainite X80 pipeline steel. Hydrogen embrittlement susceptibility was evaluated based on fracture strain in tensile testing. We conducted a thermal desorption analysis to measure the amount of tracer hydrogen corresponding to that of lattice defects. Hydrogen embrittlement susceptibility and the amount of tracer hydrogen significantly increased with decreasing crosshead speed. Additionally, a significant increase in the formation of hydrogen-enhanced strain-induced lattice defects was observed immediately before the final fracture. In contrast to hydrogen-free specimens, the fracture surface of the hydrogen-charged specimens exhibited shallower dimples without nuclei, such as secondary phase particles. These findings indicate that the presence of hydrogen enhanced the formation of lattice defects, particularly just prior to the occurrence of final fracture. This in turn enhanced the formation of shallower dimples, thereby potentially causing premature fracture of X80 pipeline steel at lower crosshead speeds.

Results from Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants - Irradiation Embrittlement of RPV Steels -

International Nuclear Information System (INIS)

Abe, Hiroaki; Onizawa, Kunio; Katsuyama, Jinya; Murakami, Kenta; Iwai, Takeo; Iwata, Tadao; Katano, Yoshio; Sekimura, Naoto; Nagai, Yasuyoshi; Toyama, Takeshi; Tamura, Satoshi

2012-01-01

As one of the NISA Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants, we have performed research on the irradiation embrittlement of reactor pressure vessel (RPV) steels, especially focusing on irradiation embrittlement on heat affected zone (HAZ) and on applications of ion beams to deduce fundamental insights irradiation-induced embrittlement. The results obtained from the project are summarized as follows. In order to obtain the technical basis to judge the necessity of surveillance specimens from HAZ, the neutron irradiation program was performed at JRR-3, JAEA. The samples were carefully designed based on the insights from finite element analysis, metallography, 3D atom probe and positron annihilation methods, and were fabricated so as to simulate both heat treatment history and microstructure for typical HAZ from as-fabricated RPV steels which also have variation of impurity levels. The fracture toughness of the unirradiated HAZ specimens was equivalent to or better than that of base metals. Irradiation embrittlement and hardening were roughly identical to those of base metals, while some of the fine-grained HAZ microstructure was susceptible to it. The probabilistic fracture mechanics analysis was applied to the structural integrity assessment taking into account the heterogeneous microstructure as well as susceptibility for irradiation embrittlement of each HAZ microstructure under the variation of welding parameter and PTS condition. It was shown that crack propagation at the fine-grained HAZ, but the discontinuous distribution of the microstructure retards the further propagation. For the precise correlation of irradiation embrittlement of RPV steels for the long term operations, accumulations of high-dose data are required. Ion beam irradiation is one of the solutions for the regime and for mechanism-based descriptions. Another interest of ours was to describe irradiation hardening and embrittlement in terms of

Backscattered Helium Spectroscopy in the Helium Ion Microscope: Principles, Resolution and Applications

NARCIS (Netherlands)

van Gastel, Raoul; Hlawacek, G.; Dutta, S.; Poelsema, Bene

2015-01-01

We demonstrate the possibilities and limitations for microstructure characterization using backscattered particles from a sharply focused helium ion beam. The interaction of helium ions with matter enables the imaging, spectroscopic characterization, as well as the nanometer scale modification of

A novel self-embrittling strippable coating for radioactive decontamination based on silicone modified styrene-acrylic emulsion

Science.gov (United States)

Wang, Jing; Wang, Jianhui; Zheng, Li; Li, Jian; Cui, Can; Lv, Linmei

2017-03-01

Silicone modified styrene-acrylic emulsion and butyl acrylate were used as a main film-forming agent and an additive respectively to synthesize a self-embrittling strippable coating. The doping mass-ratio of butyl acrylate was adjusted at 0, 5%, 10%, 15%, 20%, and the results indicated the optimized doping ratio was 10%. Ca(OH)2 was used to promote the coating film self-embrittling at a moderate doping mass-ratio of 20%. The synthesized coating’s coefficients of α and β decontamination on concrete, marble, glass and stainless steel surfaces were both greater than 85%, which indicated the synthesized coating is a promising cleaner for radioactive decontamination.

HeREF-2003: Helium Refrigeration Techniques

CERN Multimedia

2003-01-01

CERN Technical Training 2003: Learning for the LHC ! Theory, Technology, Maintenance and Control of Helium Refrigerators HeREF-2003 is a course in the framework of the 2002 Technical Training Programme, that will provide a complete introduction to Helium refrigeration, with a practical approach to theory, technology, maintenance and control of Helium refrigeration installations. Theoretical aspects and equations will be limited to a minimum. HeREF-2003 targets an audience of technicians and operators of Helium refrigeration plants at CERN, as well as physicists and engineers needing an overview of current Helium refrigeration techniques. HeREF-2003 will address, among other, issues related to component technology, installation maintenance, process control and Helium purity. A commented visit to a couple of CERN Helium refrigeration or liquefaction plants will also take place. Duration: 7 half days (4 mornings and 3 afternoons), 6-10 October, 2003 Cost per participant: 500.- CHF Language: Bilingual English...

HeREF-2003 : Helium Refrigeration Techniques

CERN Multimedia

2003-01-01

CERN Technical Training 2003: Learning for the LHC ! Theory, Technology, Maintenance and Control of Helium Refrigerators HeREF-2003 is a course in the framework of the 2002 Technical Training Programme, that will provide a complete introduction to Helium refrigeration, with a practical approach to theory, technology, maintenance and control of Helium refrigeration installations. Theoretical aspects and equations will be limited to a minimum. HeREF-2003 targets an audience of technicians and operators of Helium refrigeration plants at CERN, as well as physicists and engineers needing an overview of current Helium refrigeration techniques. HeREF-2003 will address, among other, issues related to component technology, installation maintenance, process control and Helium purity. A commented visit to a couple of CERN Helium refrigeration or liquefaction plants will also take place. • Duration: 7 half days (4 mornings and 3 afternoons), 6-10 October, 2003 • Cost per participant: 500.- CHF
...

Calculational results for radiation embrittlement of WWER pressure vessel at the Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Apostolov, T; Ilieva, K; Petrova, T [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

Determination of radiation impact on metal state in the case of WWER-440/230 is made only by calculation methods since a special sample-witness (SW) incorporation had not been implemented. In WWER-1000 reactors such SW are foreseen but their spots are high above the active core. This is why in both reactors the appliance of a calculational procedure for radiation embrittlement determination is compulsory. The authors propose such a procedure accounting for the change in critical temperature of neutron brittleness by the neutron fluence. The neutron fluence and the shift of critical embrittlement temperature have been calculated for the maximum overloaded location and for the weld metal of the Kozloduy-5 and Kozloduy-6 reactors (WWER-1000). The shift of critical temperature in weld 4 of the Units 1-4 (WWER-440) is plotted versus work cycles and compared to experimental values. 4 figs., 5 tabs.

Fabrication of poly(methyl methacrylate)-block-poly(methacrylic acid) diblock copolymer as a self-embrittling strippable coating for radioactive decontamination

International Nuclear Information System (INIS)

Liu Renlong; Zhang Huiyan; Li Yintao; Zhou Yuanlin; Zhang Quanping; Zheng Jian; Wang Shanqiang

2016-01-01

The poly(methyl methacrylate)-block-poly(methacrylic acid) diblock copolymer with different monomer compositions was synthesized via reversible addition-fragmentation chain transfer polymerization. Meanwhile, a novel self-embrittling strippable coating was prepared using the diblock copolymers, which is proposed to be used as radioactive decontamination agents without manual operation. Furthermore, the decontamination efficiencies of self-embrittling strippable coatings for radioactive contamination on glass, marble, and stainless steel surfaces were studied. (author)

Helium localization around the microscopic impurities embedded to liquid helium

International Nuclear Information System (INIS)

Gordon, E.B.; Shestakov, A.F.

2000-01-01

The structure and properties of the environment round the impurity atoms (Im) embedded in liquid helium are considered. It is shown that there are two qualitatively different types of structure of the He atom layer next to Im - attraction and repulsion structures. For the center attraction structure (strong Im-He interaction) the Im-He separation is longer than the equilibrium one for the pair Im-He potential, and the density and localization of He atoms are higher than in the bulk. It this case the He atom content in the layer, n, is almost independent of applied pressure. In the repulsion structure realized for alkaline metal atoms the Im-He separation is shorter than the equilibrium one and the density is lower than in the helium bulk. At T approx 1 K occupied are several states with different n and their energies differ only by approx 0.1 K, an increase in pressure resulting in a considerable reduction of n. The optical and EPR spectra of the atoms embedded to liquid and solid helium are interpreted on the basis of the analysis carried out. A simple model is proposed to evaluate the helium surroundings characteristics from the experimental pressure dependences of atomic line shifts in the absorption and emission spectra. The attraction structures in 3 He - 4 He mixtures are suggested to be highly enriched by 4 He atoms which the repulsion structures - by 3 He atoms. a possibility for existence of phase transitions in helium shells surrounding impurity atoms is considered

Effects of surface condition on aqueous corrosion and environmental embrittlement of iron aluminides

Energy Technology Data Exchange (ETDEWEB)

Perrin, R.L.; Buchanan, R.A. [Univ. of Tennessee, Knoxville, TN (United States)

1996-08-01

Effects of retained high-temperature surface oxides, produced during thermomechanical processing and/or heat treatment, on the aqueous-corrosion and environmental-embrittlement characteristics of Fe{sub 3}Al-based iron aluminides (FA-84, FA-129 and FAL-Mo), a FeAl-based iron aluminide (FA-385), and a disordered low-aluminum Fe-Al alloy (FAPY) were evaluated. All tests were conducted at room temperature in a mild acid-chloride solution. In cyclic-anodic-polarization testing for aqueous-corrosion behavior, the surface conditions examined were: as-received (i.e., with the retained high-temperature oxides), mechanically cleaned and chemically cleaned. For all materials, the polarization tests showed the critical pitting potentials to be significantly lower in the as-received condition than in the mechanically-cleaned and chemically-cleaned conditions. These results indicate detrimental effects of the retained high-temperature oxides in terms of increased susceptibilities to localized corrosion. In 200-hour U-bend stress-corrosion-cracking tests for environmental-embrittlement behavior, conducted at open-circuit corrosion potentials and at a hydrogen-charging potential of {minus}1500 mV (SHE), the above materials (except FA-385) were examined with retained oxides and with mechanically cleaned surfaces. At the open-circuit corrosion potentials, none of the materials in either surface condition underwent cracking. At the hydrogen-charging potential, none of the materials with retained oxides underwent cracking, but FA-84, FA-129 and FAL-Mo in the mechanically cleaned condition did undergo cracking. These results suggest beneficial effects of the retained high-temperature oxides in terms of increased resistance to environmental hydrogen embrittlement.

Test of a cryogenic helium pump

International Nuclear Information System (INIS)

Lue, J.W.; Miller, J.R.; Walstrom, P.L.; Herz, W.

1981-01-01

The design of a cryogenic helium pump for circulating liquid helium in a magnet and the design of a test loop for measuring the pump performance in terms of mass flow vs pump head at various pump speeds are described. A commercial cryogenic helium pump was tested successfully. Despite flaws in the demountable connections, the piston pump itself has performed satisfactorily. A helium pump of this type is suitable for the use of flowing supercritical helium through Internally Cooled Superconductor (ICS) magnets. It has pumped supercritical helium up to 7.5 atm with a pump head up to 2.8 atm. The maximum mass flow rate obtained was about 16 g/s. Performance of the pump was degraded at lower pumping speeds

Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C

International Nuclear Information System (INIS)

Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.; Garner, F.A.

1998-01-01

This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at ∼270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented levels of swelling reached at 335--360 C at these high neutron fluences. The failure mechanism appears to be identical to that observed at similar swelling levels in other austenitic steels irradiated in US fast reactors at 400--425 C, whereby stress-concentration between voids and nickel segregation at void surfaces predisposes the steel to an epsilon martensite transformation followed by formation of alpha martensite at crack tips. The very slow strain rate inherent in such creep tests and the relatively high helium levels may also contribute to the failure

Post-giant evolution of helium stars

International Nuclear Information System (INIS)

Schoenberner, D.

1977-01-01

Extremely hydrogen deficient stars (helium stars and R Coronae Borealis variables) are considered to be remnants of double shell source stars (of the asymptotic giant branch). The evolution of stars with a condensed C/O-core and a helium envelope is followed numerically from the red giant stage to the white dwarf domain, crossing the regions of R CrB- and helium stars (so far analyzed). They have typically masses M/M(sun) = 0.7 and luminosities log L/L(sun) = 4.1. The time for crossing the helium star domain is some 10 3 years. The corresponding times in the R CrB-region amounts up to several 10 4 years. The lower limit of the death rate of helium stars is estimated to be 4 x 10 -14 pc -3 yr -1 . This value is only a factor of ten lower than the birth rate of all non-DA white dwarfs. It is therefore possible that the helium stars are the precursors of helium rich white dwarfs. As a consequence, a significant fraction of all stars which end their lives as white dwarfs should pass through the helium star phase. (orig.) [de

Investigation of impurity-helium solid phase decomposition

International Nuclear Information System (INIS)

Boltnev, R.E.; Gordon, E.B.; Krushinskaya, I.N.; Martynenko, M.V.; Pel'menev, A.A.; Popov, E.A.; Khmelenko, V.V.; Shestakov, A.F.

1997-01-01

The element composition of the impurity-helium solid phase (IHSP), grown by injecting helium gas jet, involving Ne, Ar, Kr, and Xe atoms and N 2 molecules, into superfluid helium, has been studied. The measured stoichiometric ratios, S = N H e / N I m, are well over the values expected from the model of frozen together monolayer helium clusters. The theoretical possibility for the freezing of two layers helium clusters is justified in the context of the model of IHSP helium subsystem, filled the space between rigid impurity centers. The process of decomposition of impurity-helium (IH)-samples taken out of liquid helium in the temperature range 1,5 - 12 K and the pressure range 10-500 Torr has been studied. It is found that there are two stages of samples decomposition: a slow stage characterized by sample self cooling and a fast one accompanied by heat release. These results suggest, that the IHSP consists of two types of helium - weakly bound and strongly bound helium - that can be assigned to the second and the first coordination helium spheres, respectively, formed around heavy impurity particles. A tendency for enhancement of IHSP thermo stability with increasing the impurity mass is observed. Increase of helium vapor pressure above the sample causes the improvement of IH sample stability. Upon destruction of IH samples, containing nitrogen atoms, a thermoluminescence induced by atom recombination has been detected in the temperature region 3-4,5 K. This suggests that numerous chemical reactions may be realized in solidified helium

Further application of the cleavage fracture stress model for estimating the T{sub 0} of highly embrittled ferritic steels

Energy Technology Data Exchange (ETDEWEB)

Sreenivasan, P.R.

2016-02-15

The semi-empirical cleavage fracture stress model (CFS), based on the microscopic cleavage fracture stress, s{sub f}, for estimating the ASTM E1921 reference temperature (T{sub 0}) of ferritic steels from instrumented impact testing of unprecracked Charpy V-notch specimens is further confirmed by test results for additional steels, including steels highly embrittled by thermal aging or irradiation. In addition to the ferrite-pearlite, bainitic or tempered martensitic steels (which was examined earlier), acicular or polygonal ferrite, precipitation-strengthened or additional simulated heat affected zone steels are also evaluated. The upper limit for the applicability of the present CFS model seems to be T{sub 41J} ∝160 to 170 C or T{sub 0} or T{sub Qcfs} (T{sub 0} estimate from the present CFS model) ∝100 to 120 C. This is not a clear-cut boundary, but indicative of an area of caution where generation and evaluation of further data are required. However, the present work demonstrates the applicability of the present CFS model even to substantially embrittled steels. The earlier doubts expressed about T{sub Qcfs} becoming unduly non-conservative for highly embrittled steels has not been fully substantiated and partly arises from the necessity of modifications in the T{sub 0} evaluation itself at high degrees of embrittlement suggested in the literature.

Mixed helium-3 - helium-4 calorimeter. Very low temperature calorimetry; Calorimetre mixte a helium-3 et helium-4. Calorimetrie a tres basse temperature

Energy Technology Data Exchange (ETDEWEB)

Testard, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1966-06-01

A description is given of a double-racket calorimeter using helium-4 and helium-3 as the cryogenic fluids and making it possible to vary the temperature continuously from 0.35 K to 4.2 K. By using an electric thermal regulator together with liquid hydrogen it is possible to extend this range up to about 30 K. In the second part, a review is made of the various, methods available for measuring specific heats. The method actually used in the apparatus previously described is described in detail. The difficulties arising from the use of an exchange gas for the thermal contact have been solved by the use of adsorption pumps. (author) [French] On decrit un calorimetre a double enceinte utilisant comme fluide cryogenique l'helium-4 et l'helium-3 et permettant de varier continuement la temperature de 0,35 K a 4,2 K. L'utilisation d'un regulateur thermique electrique ainsi que celle d'hydrogene, liquide permettent d'etendre cette gamme jusqu'a 30 K environ. Dans une deuxieme partie, on passe en revue les diverses methodes de mesure des chaleurs specifiques. La methode concrete utilisee dans l'appareil precedemment decrit est exposee en detail. Les difficultes inherentes a l'utilisation de gaz d'echange comme agent de contact thermique ont ete levees par la mise en oeuvre de pompes a adsorbant. (auteur)

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

International Nuclear Information System (INIS)

Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

1990-10-01

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

Assessment of the French and US embrittlement trend curves applied to RPV materials irradiated in the BR2 materials test reactor

International Nuclear Information System (INIS)

Chaouadi, R.; Gerard, R.; Boagaerts, A.S.

2011-01-01

The irradiation embrittlement of reactor pressure vessels (RPVs) in monitored through the surveillance programs associated with predictive formulas, the so-called embrittlement trend curves. These formulas are generally empirically derived and contain the major embrittlement-inducing elements such as copper, nickel and phosphorus. There are a number of such trend curves used in various regulatory guides used in the US, France, Germany, Russia and Japan. These trend curves are often supported by surveillance data and regularly assessed in view of updated surveillance databases. With the recent worldwide move towards life extension of existing reactors above their initially-scheduled lifetime of 40 years, adequate and accurate modeling of irradiation embrittlement becomes a concern for long term operation. The aim of this work is to assess the performance of the embrittlement trend curves used in a regulatory perspective. The work presented here is limited to US and French trend curves because the reactor pressure vessels of the Belgian nuclear power plants are either Westinghouse or Framatome design. The chemical composition of the Belgian RPVs being very close to the one of the French 900 MW units, the French trend curve is used except for the Doel 1-2 units for which these curves are not applicable due to the higher copper content of the welds. In this case, the U.S. trend curves are used. The aim of this work is to evaluate the performance of the embrittlement trend curves used in a regulatory perspective to represent the experimental data obtained in the BR2 reactor. In particular, the French (FIM, FIS) and the US (Reg. Guide 1.99 Rev. 2, ASTM E900-02, EWO and EONY) formulas are of prime interest. The results obtained clearly show that the French trend curves tend to over-estimate the actual irradiation hardening while the US curves under-estimate it. Within the long term operation perspective, both over- and under-estimating are undesirable and therefore the

Evaluation of liquid metal embrittlement of SS304 by Cd and Cd-Al solutions

International Nuclear Information System (INIS)

Thomas, J.K.; Iyer, N.C.; Begley, J.A.

1992-01-01

The susceptibility of stainless steel 304 to liquid metal embrittlement (LME) by cadmium (Cd) and cadmium-aluminum (Cd-Al) solutions was examined as part of a failure evaluation for SS304-clad cadmium reactor safety rods which had been exposed to elevated temperatures. The active, or cadmium (Cd) bearing, portion of the safety rod consists of a 0.756 in. diameter aluminum allow (Al-6061) core, a 0.05 in. thick Cd layer, and a 0.042 in. thick Type 304 stainless steel cladding. The safety rod thermal tests were conducted as part of a program to define the response of reactor core components to a hypothetical LOCA for the Savannah River Site (SRS) production reactor. LME was considered as a potential failure mechanism based on the nature of the failure and susceptibility of austenitic stainless steels to embrittlement by other liquid metals

Liquid Zn assisted embrittlement of advanced high strength steels with different microstructures

Science.gov (United States)

Jung, Geunsu; Woo, In Soo; Suh, Dong Woo; Kim, Sung-Joon

2016-03-01

In the present study, liquid metal embrittlement (LME) phenomenon during high temperature deformation was investigated for 3 grades of Zn-coated high strength automotive steel sheets consisting of different phases. Hot tensile tests were conducted for each alloy to compare their LME sensitivities at temperature ranges between 600 and 900 °C with different strain rates. The results suggest that Zn embrittles all the Fe-alloy system regardless of constituent phases of the steel. As hot tensile temperature and strain rate increase, LME sensitivity increases in every alloy. Furthermore, it is observed that the critical strain, which is experimentally thought to be 0.4% of strain at temperatures over 700 °C, is needed for LME to occur. It is observed via TEM work that Zn diffuses along grain boundaries of the substrate alloy when the specimen is strained at high temperatures. When the specimen is exposed to the strain more than 0.4% at over 700 °C, the segregation level of Zn at grain boundaries seems to become critical, leading to occurrence of LME cracks.

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.

Determination of helium in beryl minerals

International Nuclear Information System (INIS)

Souza Barcellos, E. de.

1985-08-01

In order to obtain the diffusion coefficients of helium in beryl and phenacite samples at various temperatures, helium leak rates were measured in these minerals at these temperatures. Mass spectrometry (MS) was used to obtain helium leak rates and the gas flow was plotted against time. The gas quantity determined by MS was first obtained at various temperatures until no helium leak rate was detected. After that, these samples were irradiated with fast neutrons to produce helium which was measured again. This procedure was used to estimate the experimental error. The quantity of helium produced by interaction of gamma radiation with beryl minerals was theoretically calculated from the amount of thorium-232 at the neighbourhood of the samples. The quantity of helium produced in the minerals due to uranium and thorium decay was calculated using the amount of these heavy elements, and the results were compared with the amounts determined by MS. The amount of potassium-40 was determined in order to derive the quantity of argonium-40, since some workers found argonium in excess in these minerals. The quantity of helium in the beryl samples (s) was determined in the center and in the surface of the samples in order to obtain informations about the effectiveness of the Be(α, η) He reaction. Beryl and phenacite minerals were choosen in this research since they are opposite each other with respect to the helium contents. Both have beryllium in their compositon but beryl hold a large amount of helium while phenacite, in spite of having about three times more beryllium than beryl, do not hold the gas. (author) [pt

Helium localisation in tritides

International Nuclear Information System (INIS)

Flament, J.L.; Lozes, G.

1982-06-01

Study of titanium and LaNi 5 type alloys tritides lattice parameters evolution revealed that helium created by tritium decay remains in interstitial sites up to a limit material dependant concentration. Beyond this one exceeding helium precipites in voids [fr

Measurement of OH density and air-helium mixture ratio in an atmospheric-pressure helium plasma jet

International Nuclear Information System (INIS)

Yonemori, Seiya; Ono, Ryo; Nakagawa, Yusuke; Oda, Tetsuji

2012-01-01

The absolute density of OH radicals in an atmospheric-pressure helium plasma jet is measured using laser-induced fluorescence (LIF). The plasma jet is generated in room air by applying a pulsed high voltage onto a quartz tube with helium gas flow. The time-averaged OH density is 0.10 ppm near the quartz tube nozzle, decreasing away from the nozzle. OH radicals are produced from water vapour in the helium flow, which is humidified by water adsorbed on the inner surface of the helium line and the quartz tube. When helium is artificially humidified using a water bubbler, the OH density increases with humidity and reaches 2.5 ppm when the water vapour content is 200 ppm. Two-dimensional distribution of air-helium mixture ratio in the plasma jet is also measured using the decay rate of the LIF signal waveform which is determined by the quenching rate of laser-excited OH radicals. (paper)

Effect of ternary solute interaction on interfacial segregation and grain boundary embrittlement

Czech Academy of Sciences Publication Activity Database

Lejček, Pavel

2013-01-01

Roč. 48, č. 14 (2013), 4965-4972 ISSN 0022-2461 R&D Projects: GA MŠk(CZ) LM2011026; GA ČR GAP108/12/0144 Institutional research plan: CEZ:AV0Z10100520 Keywords : interfacial segregation * intergranular embrittlement * solute interaction * modeling * thermodynamics Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.305, year: 2013

Ultra-High Efficiency / Low Hydrogen Embrittlement Nanostructured Zn-Based Electrodeposits as Environmentally Benign Cd-Replacement Coatings for High Strength Steel Fasteners

Science.gov (United States)

2011-04-01

sample production for the testing of hydrogen re-embrittlement ( HRE ) (a.k.a. in-service embrittlement); (4) further optimization of plating conditions...Ni range. This could help explain the HRE performance as a nickel concentration of 15wt.% had an OCP close to that of Cd and steel, which would...ZnNi plating, including superior corrosion protection and improved HRE performance as a result of the dense fine grained microstructure. Furthermore

Resistivity studies of interstitial helium mobility in niobium

International Nuclear Information System (INIS)

Chen, C.G.; Birnbaum, H.K.; Johnson, A.B. Jr.

1979-01-01

The mobility of interstitial helium in Nb and Nb-O alloys was studied in the temperature range of 10-383 K using resistivity measurements. The helium was introduced by radioactive decay of solute tritium (approximately 1 at%). At T < 100 K the resistivity increased due to conversion of tritium trapped at oxygen interstititals to helium. The formation of helium caused a very significant resistance increase at room temperature and above. The results suggest that helium is mobile at temperatures above 295 K and that the precipitation of large helium bubbles occurs along grain boundaries. The mobile helium species may either be single interstitials or small helium clusters. The activation enthalpy for the diffusion of the mobile helium species was estimated to be about 55 kJ/mol (0.66 eV). (Auth.)

Helium-induced hardening effect in polycrystalline tungsten

Science.gov (United States)

Kong, Fanhang; Qu, Miao; Yan, Sha; Zhang, Ailin; Peng, Shixiang; Xue, Jianming; Wang, Yugang

2017-09-01

In this paper, helium induced hardening effect of tungsten was investigated. 50 keV He2+ ions at fluences vary from 5 × 1015 cm-2 to 5 × 1017 cm-2 were implanted into polycrystalline tungsten at RT to create helium bubble-rich layers near the surface. The microstructure and mechanical properties of the irradiated specimens were studied by TEM and nano-indentor. Helium bubble rich layers are formed in near surface region, and the layers become thicker with the rise of fluences. Helium bubbles in the area of helium concentration peak are found to grow up, while the bubble density is almost unchanged. Obvious hardening effect is induced by helium implantation in tungsten. Micro hardness increases rapidly with the fluence firstly, and more slowly when the fluence is above 5 × 1016 cm-2. The hardening effect of tungsten can be attributed to helium bubbles, which is found to be in agreement with the Bacon-Orowan stress formula. The growing diameter is the major factor rather than helium bubbles density (voids distance) in the process of helium implantation at fluences below 5 × 1017 cm-2.

Low helium flux from the mantle inferred from simulations of oceanic helium isotope data

Science.gov (United States)

Bianchi, Daniele; Sarmiento, Jorge L.; Gnanadesikan, Anand; Key, Robert M.; Schlosser, Peter; Newton, Robert

2010-09-01

The high 3He/ 4He isotopic ratio of oceanic helium relative to the atmosphere has long been recognized as the signature of mantle 3He outgassing from the Earth's interior. The outgassing flux of helium is frequently used to normalize estimates of chemical fluxes of elements from the solid Earth, and provides a strong constraint to models of mantle degassing. Here we use a suite of ocean general circulation models and helium isotope data obtained by the World Ocean Circulation Experiment to constrain the flux of helium from the mantle to the oceans. Our results suggest that the currently accepted flux is overestimated by a factor of 2. We show that a flux of 527 ± 102 mol year - 1 is required for ocean general circulation models that produce distributions of ocean ventilation tracers such as radiocarbon and chlorofluorocarbons that match observations. This new estimate calls for a reevaluation of the degassing fluxes of elements that are currently tied to the helium fluxes, including noble gases and carbon dioxide.

Observation of visible emission from the molecular helium ion in the afterglow of a dense helium Z-pinch plasma

International Nuclear Information System (INIS)

Tucker, J.E.; Brake, M.L.; Gilgenbach, R.M.

1986-01-01

The authors present the results of axial and radial time resolved visible emission spectroscopy from the afterglow of a dense helium Z-pinch. These results show that the visible emissions in the pinch afterglow are dominated by line emissions from molecular helium and He II. Axial spectroscopy measurements show the occurrence of several absorption bands which cannot be identified as molecular or atomic helium nor impurities from the discharge chamber materials. The authors believe that these absorption bands are attributable to the molecular helium ion which is present in the discharge. The molecular ion has been observed by others in low pressure and temperature helium discharges directly by means of mass spectrometry and indirectly by the presence of helium atoms in the 2/sup 3/S state, (the He 2/sup 3/S state is believed to result from molecular helium ion recombination). However, the molecular helium ion has not previously been observed spectroscopically

Helium Extraction from LNG End Flash

OpenAIRE

Kim, Donghoi

2014-01-01

Helium is an invaluable element as it is widely used in industry such as cryo-genics and welding due to its unique properties. However, helium shortage is expected in near future because of increasing demand and the anxiety of sup-ply. Consequently, helium production has attracted the attention of industry. The main source of He is natural gas and extracting it from LNG end-flash is considered as the most promising way of producing crude helium. Thus, many process suppliers have proposed proc...

Use of separating nozzles or ultra-centrifuges for obtaining helium from gas mixtures containing helium

International Nuclear Information System (INIS)

Reimann, T.

1987-01-01

To obtain helium from gas mixtures containing helium, particularly from natural gas, it is proposed to match the dimensions of the separation devices for a ratio of the molecular weights to be separated of 4:1 of more, which ensures a higher separation factor and therefore a smaller number of separation stages to be connected in series. The process should make reasonably priced separation of helium possible. (orig./HP) [de

Helium diffusion in nickel at high temperatures

International Nuclear Information System (INIS)

Philipps, V.

1980-09-01

Helium has been implanted at certain temperatures between 800 and 1250 0 C into single and polycrystalline Ni-samples with implantation depths between 15 and 90 μm. Simultaneously the helium reemission from the sample is measured by a mass-spectrometer. It has been shown that the time dependence of the observed reemission rate is governed by volume diffusion of the helium. Measuring this time dependence as a function of temperature the helium diffusion constant has been determined. The He-diffusion is interpreted as a interstitial diffusion hindered by thermal vacancies. Depending on the implantation depth more or less of the implanted helium remains in the sample and forms large helium bubbles. (orig./GSCH)

Neutron-induced helium implantation in GCFR cladding

International Nuclear Information System (INIS)

Yamada, H.; Poeppel, R.B.; Sevy, R.H.

1980-10-01

The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

Quantum statistics and liquid helium 3 - helum 4 mixtures

International Nuclear Information System (INIS)

Cohen, E.G.D.

1979-01-01

The behaviour of liquid helium 3-helium 4 mixtures is considered from the point of view of manifestation of quantum statistics effects in macrophysics. The Boze=Einstein statistics is shown to be of great importance for understanding superfluid helium-4 properties whereas the Fermi-Dirac statistics is of importance for understanding helium-3 properties. Without taking into consideration the interaction between the helium atoms it is impossible to understand the basic properties of liquid helium 33 - helium 4 mixtures at constant pressure. Proposed is a simple model of the liquid helium 3-helium 4 mixture, namely the binary mixture consisting of solid spheres of two types subjecting to the Fermi-Dirac and Bose-Einstein statistics relatively. This model predicts correctly the most surprising peculiarities of phase diagrams of concentration dependence on temperature for helium solutions. In particular, the helium 4 Bose-Einstein statistics is responsible for the phase lamination of helium solutions at low temperatures. It starts in the peculiar critical point. The helium 4 Fermi-Dirac statistics results in incomplete phase lamination close to the absolute zero temperatures, that permits operation of a powerful cooling facility, namely refrigerating machine on helium solution

Effect of solute interaction on interfacial and grain boundary embrittlement in binary alloys

Czech Academy of Sciences Publication Activity Database

Lejček, Pavel

2013-01-01

Roč. 48, č. 6 (2013), 2574-2580 ISSN 0022-2461 R&D Projects: GA ČR GAP108/12/0144 Institutional research plan: CEZ:AV0Z10100520 Keywords : interfacial segregation * grain boundary embrittlement * binary interaction * modeling * thermodynamics Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.305, year: 2013

High Efficiency Regenerative Helium Compressor, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Helium plays several critical rolls in spacecraft propulsion. High pressure helium is commonly used to pressurize propellant fuel tanks. Helium cryocoolers can be...

Seismological measurement of solar helium abundance

International Nuclear Information System (INIS)

Vorontsov, S.V.; Pamyatnykh, A.A.

1991-01-01

The internal structure and evolution of the Sun depends on its chemical composition, particularly the helium abundance. In addition, the helium abundance in the solar envelope is thought to represent the protosolar value, making it a datum of cosmological significance. Spectroscopic measurements of the helium abundance are uncertain, and the most reliable estimates until now have come from the calibration of solar evolutionary models. The frequencies of solar acoustic oscillations are sensitive, however, to the behaviour of the speed of sound in the Sun's helium ionization zone, which allows a helioseismological determination of the helium abundance. Sound-speed inversion of helioseismological data can be used for this purpose, but precise frequency measurements of high-degree oscillation modes are needed. Here we describe a new approach based on an analysis of the phase shift of acoustic waves of intermediate-degree modes. From the accurate intermediate-mode data now available, we obtain a helium mass fraction Y=0.25±0.01 in the solar convection zone, significantly smaller than the value Y=0.27-0.29 predicted by recent solar evolutionary models. The discrepancy indicates either that initial helium abundance was reduced in the envelope by downward diffusion or that the protosolar value was lower than currently accepted. (author)

Review of Membranes for Helium Separation and Purification

Directory of Open Access Journals (Sweden)

Colin A. Scholes

2017-02-01

Full Text Available Membrane gas separation has potential for the recovery and purification of helium, because the majority of membranes have selectivity for helium. This review reports on the current state of the research and patent literature for membranes undertaking helium separation. This includes direct recovery from natural gas, as an ancillary stage in natural gas processing, as well as niche applications where helium recycling has potential. A review of the available polymeric and inorganic membranes for helium separation is provided. Commercial gas separation membranes in comparable gas industries are discussed in terms of their potential in helium separation. Also presented are the various membrane process designs patented for the recovery and purification of helium from various sources, as these demonstrate that it is viable to separate helium through currently available polymeric membranes. This review places a particular focus on those processes where membranes are combined in series with another separation technology, commonly pressure swing adsorption. These combined processes have the most potential for membranes to produce a high purity helium product. The review demonstrates that membrane gas separation is technically feasible for helium recovery and purification, though membranes are currently only applied in niche applications focused on reusing helium rather than separation from natural sources.

Separation of compressor oil from helium

International Nuclear Information System (INIS)

Strauss, R.; Perrotta, K.A.

1982-01-01

Compression of helium by an oil-sealed rorary screw compressor entrains as much as 4000 parts per million by weight of liquid and vapor oil impurities in the gas. The reduction below about 0.1 ppm for cryogenic applications is discussed. Oil seperation equipment designed for compressed air must be modified significantly to produce the desired results with helium. The main differences between air and helium filtration are described. A description of the coalescers is given with the continuous coalescing of liquid mist from air or other gas illustrated. Oil vapor in helium is discussed in terms of typical compressor oils, experimental procedure for measuring oil vapor concentration, measured volatile hydrocarbons in the lubricants, and calculated concentration of oil vapor in Helium. Liquid oil contamination in helium gas can be reduced well below 0.1 ppm by a properly designed multiple state coalescing filter system containing graded efficiency filter elements. The oil vapor problem is best attached by efficiently treating the oil to remove most of the colatiles before charging the compressor

Effects of strain rate, stress condition and environment on iodine embrittlement of Ziracloy-2

International Nuclear Information System (INIS)

Une, K.

1979-01-01

Iodine stress corrosion cracking (SCC) susceptibility of Zircaloy became higher with decreasing strain rate. Critical strain rate, below which high SCC severity was observed, substantially depended on Zircaloy stress condition. This strain rate (7 x 10 -3 min -1 ) under plane strain condition was about 3.5 times as fast as that (2 x 10 -3 min -1 ) under uniaxial condition. The maximum iodine embrittlement in Zircaloy was found in stress ratio α (axial/tangential stress) range of 0.5 to 0.7. No embrittlement occurred at α = infinity because of its texture effect. The SCC fracture stresses were about 39 kg/mm 2 for unirradiated and stress-relieved material, and about 34 kg/mm 2 for recrystallized material, whose ratios to yield strength of each material were 0.8 and 1.2. Impurity gases of oxygen and moisture in the iodine had the effects of reducing Zircaloy SCC susceptibility. Stress-relieved material was more sensitive to environmental impurities than recrystallized material

Low-temperature embrittlement and fracture of metals with different crystal lattices – Dislocation mechanisms

Directory of Open Access Journals (Sweden)

V.M. Chernov

2016-12-01

Full Text Available The state of a low-temperature embrittlement (cold brittleness and dislocation mechanisms for formation of the temperature of a ductile-brittle transition and brittle fracture of metals (mono- and polycrystals with various crystal lattices (BCC, FCC, HCP are considered. The conditions for their formation connected with a stress-deformed state and strength (low temperature yield strength as well as the fracture breaking stress and mobility of dislocations in the top of a crack of the fractured metal are determined. These conditions can be met for BCC and some HCP metals in the initial state (without irradiation and after a low-temperature damaging (neutron irradiation. These conditions are not met for FCC and many HCP metals. In the process of the damaging (neutron irradiation such conditions are not met also and the state of low-temperature embrittlement of metals is absent (suppressed due to arising various radiation dynamic processes, which increase the mobility of dislocations and worsen the strength characteristics.

Nuclear fuel rod helium leak inspection apparatus and method

International Nuclear Information System (INIS)

Ahmed, H.J.

1991-01-01

This patent describes an inspection apparatus for testing nuclear fuel rods for helium leaks. It comprises a test chamber being openable and closable for receiving at least one nuclear fuel rod; means separate from the fuel rod for supplying helium and constantly leaking helium at a predetermined known positive value into the test chamber to constantly provide an atmosphere of helium at the predetermined known positive value in the test chamber; and means for sampling the atmosphere within the chamber and measuring the helium in the atmosphere such that a measured helium value below a preset minimum helium value substantially equal to the predetermined known positive value of the atmosphere of helium being constantly provided in the test chamber indicates a malfunction in the inspection apparatus, above a preset maximum helium value greater than the predetermined known positive in the test chamber indicates the existence of a helium leak from the fuel rod, or between the preset minimum and maximum helium values indicates the absence of a helium leak from the fuel rod

Hydrogen embrittlement property of a 1700-MPa-class ultrahigh-strength tempered martensitic steel

Energy Technology Data Exchange (ETDEWEB)

Li Songjie; Zhang Boping [School of Materials Science and Engineering, University of Science and Technology Beijing, No. 30 Xueyuan Road, Hidian Zone, Beijing 100083 (China); Akiyama, Eiji; Yuuji, Kimura; Tsuzaki, Kaneaki [Structural Metals Center, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Uno, Nobuyoshi, E-mail: AKIYAMA.Eiji@nims.go.j [Nippon Steel and Sumikin Metal Products Co, Ltd, SA Bldg., 17-12 Kiba 2-chome, Koto-ku, Tokyo (Japan)

2010-04-15

The hydrogen embrittlement property of a prototype 1700-MPa-class ultrahigh-strength steel (NIMS17) containing hydrogen traps was evaluated using a slow strain rate test (SSRT) after cathodic hydrogen precharging, cyclic corrosion test (CCT) and atmospheric exposure. The hydrogen content in a fractured specimen was measured after SSRT by thermal desorption spectroscopy (TDS). The relationship between fracture stress and hydrogen content for the hydrogen-precharged specimens showed that the fracture stress of NIMS17 steel was higher, at a given hydrogen content, than that of conventional AISI 4135 steels with tensile strengths of 1300 and 1500 MPa. This suggests better resistance of NIMS17 steel to hydrogen embrittlement. However, hydrogen uptake to NIMS17 steel under CCT and atmospheric exposure decreased the fracture stress. This is because of the stronger hydrogen uptake to the steel containing hydrogen traps than to the AISI 4135 steels. Although NIMS17 steel has a higher strength level than AISI 4135 steel with a tensile strength of 1500 MPa, the decrease in fracture stress is similar between these steels.

Hydrogen embrittlement property of a 1700-MPa-class ultrahigh-strength tempered martensitic steel

Directory of Open Access Journals (Sweden)

Songjie Li, Eiji Akiyama, Kimura Yuuji, Kaneaki Tsuzaki, Nobuyoshi Uno and Boping Zhang

2010-01-01

Full Text Available The hydrogen embrittlement property of a prototype 1700-MPa-class ultrahigh-strength steel (NIMS17 containing hydrogen traps was evaluated using a slow strain rate test (SSRT after cathodic hydrogen precharging, cyclic corrosion test (CCT and atmospheric exposure. The hydrogen content in a fractured specimen was measured after SSRT by thermal desorption spectroscopy (TDS. The relationship between fracture stress and hydrogen content for the hydrogen-precharged specimens showed that the fracture stress of NIMS17 steel was higher, at a given hydrogen content, than that of conventional AISI 4135 steels with tensile strengths of 1300 and 1500 MPa. This suggests better resistance of NIMS17 steel to hydrogen embrittlement. However, hydrogen uptake to NIMS17 steel under CCT and atmospheric exposure decreased the fracture stress. This is because of the stronger hydrogen uptake to the steel containing hydrogen traps than to the AISI 4135 steels. Although NIMS17 steel has a higher strength level than AISI 4135 steel with a tensile strength of 1500 MPa, the decrease in fracture stress is similar between these steels.

Hydrogen embrittlement property of a 1700-MPa-class ultrahigh-strength tempered martensitic steel

International Nuclear Information System (INIS)

Li Songjie; Zhang Boping; Akiyama, Eiji; Yuuji, Kimura; Tsuzaki, Kaneaki; Uno, Nobuyoshi

2010-01-01

The hydrogen embrittlement property of a prototype 1700-MPa-class ultrahigh-strength steel (NIMS17) containing hydrogen traps was evaluated using a slow strain rate test (SSRT) after cathodic hydrogen precharging, cyclic corrosion test (CCT) and atmospheric exposure. The hydrogen content in a fractured specimen was measured after SSRT by thermal desorption spectroscopy (TDS). The relationship between fracture stress and hydrogen content for the hydrogen-precharged specimens showed that the fracture stress of NIMS17 steel was higher, at a given hydrogen content, than that of conventional AISI 4135 steels with tensile strengths of 1300 and 1500 MPa. This suggests better resistance of NIMS17 steel to hydrogen embrittlement. However, hydrogen uptake to NIMS17 steel under CCT and atmospheric exposure decreased the fracture stress. This is because of the stronger hydrogen uptake to the steel containing hydrogen traps than to the AISI 4135 steels. Although NIMS17 steel has a higher strength level than AISI 4135 steel with a tensile strength of 1500 MPa, the decrease in fracture stress is similar between these steels.

Helium behaviour in implanted boron carbide

Directory of Open Access Journals (Sweden)

Motte Vianney

2015-01-01

Full Text Available When boron carbide is used as a neutron absorber in nuclear power plants, large quantities of helium are produced. To simulate the gas behaviour, helium implantations were carried out in boron carbide. The samples were then annealed up to 1500 °C in order to observe the influence of temperature and duration of annealing. The determination of the helium diffusion coefficient was carried out using the 3He(d,p4He nuclear reaction (NRA method. From the evolution of the width of implanted 3He helium profiles (fluence 1 × 1015/cm2, 3 MeV corresponding to a maximum helium concentration of about 1020/cm3 as a function of annealing temperatures, an Arrhenius diagram was plotted and an apparent diffusion coefficient was deduced (Ea = 0.52 ± 0.11 eV/atom. The dynamic of helium clusters was observed by transmission electron microscopy (TEM of samples implanted with 1.5 × 1016/cm2, 2.8 to 3 MeV 4He ions, leading to an implanted slab about 1 μm wide with a maximum helium concentration of about 1021/cm3. After annealing at 900 °C and 1100 °C, small (5–20 nm flat oriented bubbles appeared in the grain, then at the grain boundaries. At 1500 °C, due to long-range diffusion, intra-granular bubbles were no longer observed; helium segregates at the grain boundaries, either as bubbles or inducing grain boundaries opening.

Helium storage and control system for the PBMR

International Nuclear Information System (INIS)

Verkerk, E.C.

1997-01-01

The power conversion unit will convert the heat energy in the reactor core to electrical power. The direct-closed cycle recuperated Brayton Cycle employed for this concept consists of a primary helium cycle with helium powered turbo compressors and a power turbine. The helium is actively cooled with water before the compression stages. A recuperator is used to preheat the helium before entering the core. The start of the direct cycle is initiated by a mass flow from the helium inventory and control system via a jet pump. When the PBMR is connected to the grid, changes in power demand can be followed by changing the helium flow and pressure inside the primary loop. Small rapid adjustments can be performed without changing the helium inventory of the primary loop. The stator blade settings on the turbines and compressors are adjustable and it is possible to bypass reactor and turbine. This temporarily reduces the efficiency at which the power conversion unit is operating. Larger or long term adjustments require storage or addition of helium in order to maintain a sufficient level of efficiency in the power conversion unit. The helium will be temporarily stored in high pressure tanks. After a rise in power demand it will be injected back into the system. Some possibilities how to store the helium are presented in this paper. The change of helium inventory will cause transients in the primary helium loop in order to acquire the desired power level. At this stage, it seems that the change of helium inventory does not strongly effect the stability of the power conversion unit. (author)

Perspective on present and future alloy development efforts on austenitic stainless steels for fusion application

International Nuclear Information System (INIS)

Maziasz, P.J.

1984-01-01

The purpose of this paper is to address important questions concerning how to effect further alloy development of austenitic stainless steels for resistance, and to what extent the behavior of other properties under irradiation, such as strength/embrittlement, fatigue/irradiation creep, corrosion (under irradiation), and radiation-induced activation must be influenced. To summarize current understanding, helium has been found to have major effects on swelling and embrittlement, but several metallurgical avenues are available for significant improvement relative to type 316 stainless steel. Studies on fatigue and irradiation creep, particularly including helium effects, are preliminary but have yet to reveal engineering problems requiring additional alloy development remedies. The effects of irradiation on corrosion behavior are unknown, but higher alloy nickel contents make thermal corrosion in lithium worse. 67 refs

Helium behaviour in aluminium under hydrostatic pressure

International Nuclear Information System (INIS)

Sokurskij, Yu.N.; Tebus, V.N.; Zudilin, V.A.; Tumanova, G.M.

1989-01-01

Effect of hydrostatic compression on equilibrium helium bubbles in low aluminium-lithium alloy irradiated in reactor at 570 K is investigated. Measurements of hydrostatic density and electron-microscopic investigations have shown, that application of up to 2 GPa pressure reduces equilibrium size of helium bubbles and reduces helium swelling. Kinetics and thermodynamics of the process are considered with application of 'rigid sphere' equation which describes helium state in bubbles

Preliminary Overview of a Helium Cooling System for the Secondary Helium Loop in VHTR-based SI Hydrogen Production Facilities

Energy Technology Data Exchange (ETDEWEB)

Shin, Youngjoon; Cho, Mintaek; Kim, Dahee; Lee, Taehoon; Lee, Kiyoung; Kim, Yongwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Nuclear hydrogen production facilities consist of a very high temperature gas-cooled nuclear reactor (VHTR) system, intermediate heat exchanger (IHX) system, and a sulfur-iodine (SI) thermochemical process. This study focuses on the coupling system between the IHX system and SI thermochemical process. To prevent the propagation of the thermal disturbance owing to the abnormal operation of the SI process components from the IHX system to the VHTR system, a helium cooling system for the secondary helium of the IHX is required. In this paper, the helium cooling system has been studied. The temperature fluctuation of the secondary helium owing to the abnormal operation of the SI process was then calculated based on the proposed coupling system model. Finally, the preliminary conceptual design of the helium cooling system with a steam generator and forced-draft air-cooled heat exchanger to mitigate the thermal disturbance has been carried out. A conceptual flow diagram of a helium cooling system between the IHX and SI thermochemical processes in VHTR-based SI hydrogen production facilities has been proposed. A helium cooling system for the secondary helium of the IHX in this flow diagram prevents the propagation of the thermal disturbance from the IHX system to the VHTR system, owing to the abnormal operation of the SI process components. As a result of a dynamic simulation to anticipate the fluctuations of the secondary helium temperature owing to the abnormal operation of the SI process components with a hydrogen production rate of 60 mol·H{sub 2}/s, it is recommended that the maximum helium cooling capacity to recover the normal operation temperature of 450 .deg. C is 31,933.4 kJ/s. To satisfy this helium cooling capacity, a U-type steam generator, which has a heat transfer area of 12 m{sup 2}, and a forced-draft air-cooled condenser, which has a heat transfer area of 12,388.67 m{sup 2}, are required for the secondary helium cooling system.

ACPD detection and evaluation of 475 °C embrittlement of aged 2507 super duplex stainless steels

Science.gov (United States)

Gutiérrez-Vargas, Gildardo; López, Víctor H.; Carreón, Héctor; Kim, Jin-Yeon; Ruiz, Alberto

2017-02-01

An investigation to evaluate embrittlement of thermally aged 2507 super duplex stainless steel (SDSS) by means of an accurate measurement of the electric conductivity using an alternating current potential drop (ACPD) probe is conducted. Samples were aged for different periods up to 300 h at 475 °C. Results obtained from the ACPD measurements show appreciable increases in electric conductivity of samples with prolonged exposure to this temperature. In addition, the hardness of the samples increases significantly for long holding times, resulting in an embrittlement of the SDSS. These results are also supported by other data from sample-based laboratory techniques, i.e. microhardness and microscopy results which provide more direct evidences of the sensitization. This paper, therefore, demonstrates the feasibility of using the ACPD probe in field applications.

Nucleation path of helium bubbles in metals during irradiation

International Nuclear Information System (INIS)

Morishita, Kazunori

2008-01-01

Thermodynamical formalization is made for description of the nucleation and growth of helium bubbles in metals during irradiation. The proposed formalization is available or evaluating both microstructural changes in fusion first wall materials where helium is produced by (n, α) nuclear transmutation reactions, and those in fusion divertor materials where helium particles with low energy are directly implanted. Calculated nucleation barrier is significantly reduced by the presence of helium, showing that a helium bubble with an appropriate number of helium atoms depending on bubble size can nucleate without any large nucleation barriers, even at a condition where an empty void has very large nucleation barrier without helium. With the proposed thermodynamical formalization, the nucleation and growth process of helium bubbles in iron during irradiation is simulated by the kinetic Monte-Carlo (KMC) technique. It shows the nucleation path of a helium bubble on the (N He , N V ) space as functions of temperatures and the concentration of helium in the matrix, where N He and N V are the number of helium atoms and vacancies in the helium bubble, respectively. Bubble growth rates depend on the nucleation path and suggest that two different mechanisms operate for bubble growth: one is controlled by vacancy diffusion and the other is controlled by interstitial helium diffusion. (author)

Helium mobility in advanced nuclear ceramics

International Nuclear Information System (INIS)

Agarwal, Shradha

2014-01-01

The main goal of this work is to improve our knowledge on the mechanisms able to drive the helium behaviour in transition metal carbides and nitrides submitted to thermal annealing or ion irradiation. TiC, TiN and ZrC polycrystals were implanted with 3 MeV 3 He ions at room temperature in the fluence range 2 * 10 15 et 6 * 10 16 cm -2 . Some of them have been pre-irradiated with self-ions (14 MeV Ti or Zr). Fully controlled thermal annealing tests were subsequently carried out in the temperature range 1000 - 1600 C for two hours. The evolution of the helium depth distribution in function of implantation dose, temperature and pre-irradiation dose was measured thanks to the deuteron-induced nuclear reaction 3 He(d, p 0 ) 4 He between 900 keV and 1.8 MeV. The microstructure of implanted and annealed samples was investigated by transmission electron microscopy on thin foils prepared using the FIB technique. Additional characterization tools, as X-ray diffraction and Raman microspectrometry, have been also applied in order to obtain complementary information. Among the most relevant results obtained, the following have to be outlined: - double-peak helium depth profile was measured on as implanted sample for the three compounds. The first peak is located near the end of range and includes the major part of helium, a second peak located close to the surface corresponds to the helium atoms trapped by the native vacancies; - the helium retention capacity in transition metal carbides and nitrides submitted to fully controlled thermal treatments varies according to ZrC 0.92 ≤ TiC 0.96 ≤ TiN 0.96 ; - whatever the investigated material, a self-ion-induced pre-damaging does not modify the initial helium profile extent. The influence of the post-implantation thermal treatment remains preponderant in any case; - the apparent diffusion coefficient of helium is in the range 4 * 10 -18 - 2 * 10 -17 m 2 s -1 in TiC0.96 and 3.5 * 10 -19 - 5.3 * 10 -18 m 2 s -1 in TiN 0.96 between

Measurement of helium production cross sections of iron for d-T neutrons by helium accumulation method

Energy Technology Data Exchange (ETDEWEB)

Takao, Yoshiyuki; Kanda, Yukinori; Nagae, Koji; Fujimoto, Toshihiro [Kyushu Univ., Fukuoka (Japan); Ikeda, Yujiro

1997-03-01

Helium production cross sections of Iron were measured by helium accumulation method for neutron energies from 13.5 to 14.9 MeV. Iron samples were irradiated with FNS, an intense d-T neutron source of JAERI. As the neutron energy varies according to the emission angle at the neutron source, the samples were set around the neutron source and were irradiated by neutrons of different energy depending on each sample position. The amount of helium produced in a sample was measured by Helium Atoms Measurement System at Kyushu University. The results of this work are in good agreement with other experimental data in the literature and also compared with the evaluated values in JENDL-3. (author)

High-temperature helium-loop facility

International Nuclear Information System (INIS)

Tokarz, R.D.

1981-09-01

The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100 0 F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system. The experimental capabilities and test conand presents the results that have been obtained. The study has been conducted using a four-phase approach. The first phase develops the solution to the steady-state radon-diffusion equation in one-dimensieered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent f water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Uranium concentrations in the sediments which were above detection limits ranged from 0.10 t 51.2 ppM. The mean of the logarithms of the uranium concentrations was 0.53. A group of high uranium concentrations occurs near the junctions of quadrangles AB, AC, BB, a 200 mK. In case 2), x-ray studies of isotopic phase separation in 3 He-- 4 He bcc solids were carried out by B. A. Fraass

Effect of helium on void swelling in vanadium

International Nuclear Information System (INIS)

Brimhall, J.L.; Simonen, E.P.

1975-01-01

Little difference in void microstructural swelling of vanadium is observed when helium is injected simultaneously with a 46- or 5-MeV nickel beam as compared to no helium injection, at least at high dose rates. At lower dose rates, a strong helium effect is seen when the helium is injected prior to heavy ion bombardment. The effect of the helium is shown to be a strong function of the overall displacement damage rate

Vanadium alloy membranes for high hydrogen permeability and suppressed hydrogen embrittlement

International Nuclear Information System (INIS)

Kim, Kwang Hee; Park, Hyeon Cheol; Lee, Jaeho; Cho, Eunseog; Lee, Sang Mock

2013-01-01

The structural properties and hydrogen permeation characteristics of ternary vanadium–iron–aluminum (V–Fe–Al) alloy were investigated. To achieve not only high hydrogen permeability but also strong resistance to hydrogen embrittlement, the alloy composition was modulated to show high hydrogen diffusivity but reduced hydrogen solubility. We demonstrated that matching the lattice constant to the value of pure V by co-alloying lattice-contracting and lattice-expanding elements was quite effective in maintaining high hydrogen diffusivity of pure V

Clustering of Helium Atoms at a ½

NARCIS (Netherlands)

Berg, F. v.d.; Heugten, W. v.; Caspers, L.M.; Veen, A. v.; Hosson, J.Th.M. de

1977-01-01

Atomistic calculations on a ½<111>{110} edge dislocation show a restricted tendency of clustering of helium atom along this dislocation. Clusters with up to 4 helium atoms have been studied. A cluster with 3 helium proved to be most stable.

Effect of helium on void formation in nickel

International Nuclear Information System (INIS)

Brimhall, J.L.; Simonen, E.P.

1977-01-01

This study examines the influence of helium on void formation in self-ion irradiated nickel. Helium was injected either simultaneously with, or prior to, the self-ion bombardment. The void microstructure was characterized as a function of helium deposition rate and the total heavy-ion dose. In particular, at 575 0 C and 5 X 10 -3 displacements per atom per second the void density is found to be proportional to the helium deposition rate. The dose dependence of swelling is initially dominated by helium driven nucleation. The void density rapidly saturates after which swelling continues with increasing dose only from void growth. It is concluded that helium promotes void nucleation in nickel with either helium implantation technique, pre-injection or simultaneous injection. Qualitative differences, however, are recognized. (Auth.)

The installation of helium auxiliary systems in HTGR

International Nuclear Information System (INIS)

Qin Zhenya; Fu Xiaodong

1993-01-01

The inert gas Helium was chosen as reactor coolant in high temperature gas coolant reactor, therefore a set of Special and uncomplex helium auxiliary systems will be installed, the safe operation of HTR-10 can be safeguarded. It does not effect the inherent safety of HTR-10 MW if any one of all those systems were damaged during operation condition. This article introduces the design function and the system principle of all helium auxiliary systems to be installed in HTR-10. Those systems include: helium purification and its regeneration system, helium supply and storage system, pressure control and release system of primary system, dump system for helium auxiliary system and fuel handling, gaseous waste storage system, water extraction system for helium auxiliary systems and evacuation system for primary system

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

Fracture analysis of HFIR beam tube caused by radiation embrittlement

International Nuclear Information System (INIS)

Chang, S.J.

1994-01-01

With an attempt to estimate the neutron beam tube embrittlement condition for the Oak Ridge High Flux Isotope Reactor (HFIR), fracture mechanics calculations are carried out in this paper. The analysis provides some numerical result on how the tube has been structurally weakened. In this calculation, a lateral impact force is assumed. Numerical result is obtained on how much the critical crack size should be reduced if the beam tube has been subjected to an extended period of irradiation. It is also calculated that buckling strength of the tube is increased, not decreased, with irradiation

Preventing the embrittling by hydrogen when galvanizing high-grade steel

Energy Technology Data Exchange (ETDEWEB)

Paatsch, W.

1987-09-01

Galvanic precipitation of a double layer consisting of a dull nickel layer overlaid with a brilliant zinc layer on low-alloyed high-strength steel grades leads to the forming of zinc-nickel alloy layers during the subsequent heat treatment. According to traction tests carried out on high-strength steel grades, as well as to hydrogen permeability tests, this process prevents embrittling by hydrogen which might be caused by galvanic process sequences - and creates a diffusion block at the same time. The alloy layers have an excellent corrosion resistance and temperature stability.

Thirty years of screw compressors for helium; Dreissig Jahre Schraubenkompressoren fuer Helium

Energy Technology Data Exchange (ETDEWEB)

Wahl, H. [Kaeser Kompressoren GmbH, Coburg (Germany). Technisches Buero/Auftragskonstruktion

2007-07-01

KAESER helium compressors, as well as their other industrial compressors, will be further developed with the intention to improve the availability and reliability of helium liquefaction systems. Further improvement of compressor and control system efficiency will ensure a low and sustainable operating cost. Fast supply of replacement parts with several years of warranty is ensured by a world-wide distribution system and is also worked on continuously. (orig.)

Commercial helium reserves, continental rifting and volcanism

Science.gov (United States)

Ballentine, C. J.; Barry, P. H.; Hillegonds, D.; Fontijn, K.; Bluett, J.; Abraham-James, T.; Danabalan, D.; Gluyas, J.; Brennwald, M. S.; Pluess, B.; Seneshens, D.; Sherwood Lollar, B.

2017-12-01

Helium has many industrial applications, but notably provides the unique cooling medium for superconducting magnets in medical MRI scanners and high energy beam lines. In 2013 the global supply chainfailed to meet demand causing significant concern - the `Liquid Helium Crisis' [1]. The 2017 closure of Quatar borders, a major helium supplier, is likely to further disrupt helium supply, and accentuates the urgent need to diversify supply. Helium is found in very few natural gas reservoirs that have focused 4He produced by the dispersed decay (a-particle) of U and Th in the crust. We show here, using the example of the Rukwa section of the Tanzanian East African Rift, how continental rifting and local volcanism provides the combination of processes required to generate helium reserves. The ancient continental crust provides the source of 4He. Rifting and associated magmatism provides the tectonic and thermal mechanism to mobilise deep fluid circulation, focusing flow to the near surface along major basement faults. Helium-rich springs in the Tanzanian Great Rift Valley were first identified in the 1950's[2]. The isotopic compositions and major element chemistry of the gases from springs and seeps are consistent with their release from the crystalline basement during rifting [3]. Within the Rukwa Rift Valley, helium seeps occur in the vicinity of trapping structures that have the potential to store significant reserves of helium [3]. Soil gas surveys over 6 prospective trapping structures (1m depth, n=1486) show helium anomalies in 5 out of the 6 at levels similar to those observed over a known helium-rich gas reservoir at 1200m depth (7% He - Harley Dome, Utah). Detailed macroseep gas compositions collected over two days (n=17) at one site allows us to distinguish shallow gas contributions and shows the deep gas to contain between 8-10% helium, significantly increasing resource estimates based on uncorrected values (1.8-4.2%)[2,3]. The remainder of the deep gas is

Mass spectrometric analysis of helium in stainless steel

International Nuclear Information System (INIS)

Isagawa, Hiroto; Wada, Yukio; Asakura, Yoshiro; Tsuji, Nobuo; Sato, Hitoshi; Tsutsumi, Kenichi

1974-01-01

Vacuum fusion mass-spectrometry was adopted for the analysis of helium in stainless steel. Samples were heated in a vacuum crucible, and helium in the samples was extracted and collected into a reservoir tank. The gas was then introduced through an orifice into a mass spectrometer, where the amount of helium was determined. The maspeq 070 quadrupole type mass spectrometer made by Shimazu Seisakusho, Ltd. was used. The resolving power was 150, and the mass range of the apparatus was 0-150. The determination limit of helium was about 2 x 10 -3 μg when standard helium gas was analyzed, and was about 10 -2 μg when the helium in stainless steel was analyzed. The relative standard deviation of helium intensity in repetitive measurement was about 2% in the amount of helium of 0.05 μg. Helium was injected into stainless steel by means of alpha particle irradiation with a cyclotron. The amount of helium in stainless steel was then determined. The energy of alpha particles was 34 MeV, and the beam area was 10 mm x 10 mm. The experimental data were higher than the expected value in one case, and were lower in the other case. This difference was attributable to the fluctuation of alpha particle beam, misplacement of sample plates, and unevenness of the alpha beam. (Fukutomi, T.)

Alloys having improved resistance to hydrogen embrittlement

International Nuclear Information System (INIS)

Kane, R.D.; Greer, J.B.; Jacobs, D.F.; Berkowitz, B.J.

1983-01-01

The invention involves a process of improving the hydrogen embrittlement resistance of a cold-worked high yield strength nickel/cobalt base alloy containing chromium, and molybdenum and/or tungsten and having individual elemental impurity concentrations as measured by Auger spectroscopy at the crystallographic boundaries of up to about 1 Atomic percent. These elemental impurities are capable of becoming active and mobile at a temperature less than the recrystallization temperature of the alloy. The process involves heat treating the alloy at a temperature above 1300 degrees F but below the temperature of recrystallization for a time of from 1/4 to 100 hours. This is sufficient to effect a reduction in the level of the elemental impurities at the crystallographic boundaries to the range of less than 0.5 Atomic percent without causing an appreciable decrease in yield strength

Helium trapping in aluminum and sintered aluminum powders

International Nuclear Information System (INIS)

Das, S.K.; Kaminsky, M.; Rossing, T.

1975-01-01

The surface erosion of annealed aluminum and of sintered aluminum powder (SAP) due to blistering from implantation of 100-keV 4 He + ions at room temperature has been investigated. A substantial reduction in the blistering erosion rate in SAP was observed from that in pure annealed aluminum. In order to determine whether the observed reduction in blistering is due to enhanced helium trapping or due to helium released, the implanted helium profiles in annealed aluminum and in SAP have been studied by Rutherford backscattering. The results show that more helium is trapped in SAP than in aluminum for identical irradiation conditions. The observed reduction in erosion from helium blistering in SAP is more likely due to the dispersion of trapped helium at the large Al-Al 2 O 3 interfaces and at the large grain boundaries in SAP than to helium release

Helium cooling of fusion reactors

International Nuclear Information System (INIS)

Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

1994-01-01

On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

Helium induced fine structure in the electronic spectra of anthracene derivatives doped into superfluid helium nanodroplets

International Nuclear Information System (INIS)

Pentlehner, D.; Slenczka, A.

2015-01-01

Electronic spectra of organic molecules doped into superfluid helium nanodroplets show characteristic features induced by the helium environment. Besides a solvent induced shift of the electronic transition frequency, in many cases, a spectral fine structure can be resolved for electronic and vibronic transitions which goes beyond the expected feature of a zero phonon line accompanied by a phonon wing as known from matrix isolation spectroscopy. The spectral shape of the zero phonon line and the helium induced phonon wing depends strongly on the dopant species. Phonon wings, for example, are reported ranging from single or multiple sharp transitions to broad (Δν > 100 cm −1 ) diffuse signals. Despite the large number of example spectra in the literature, a quantitative understanding of the helium induced fine structure of the zero phonon line and the phonon wing is missing. Our approach is a systematic investigation of related molecular compounds, which may help to shed light on this key feature of microsolvation in superfluid helium droplets. This paper is part of a comparative study of the helium induced fine structure observed in electronic spectra of anthracene derivatives with particular emphasis on a spectrally sharp multiplet splitting at the electronic origin. In addition to previously discussed species, 9-cyanoanthracene and 9-chloroanthracene will be presented in this study for the first time

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)

The future of helium as a natural resource

CERN Document Server

Glowacki, Bartek A; Nuttall, William J

2012-01-01

The book reveals the changing dynamics of the helium industry on both the supply-side and the demand-side. The helium industry has a long-term future and this important gas will have a role to play for many decades to come. Major new users of helium are expected to enter the market, especially in nuclear energy (both fission and fusion). Prices and volumes supplied and expected to rise and this will prompt greater efforts towards the development of new helium sources and helium conservation and recycling.

Electronic properties of physisorbed helium

International Nuclear Information System (INIS)

Kossler, Sarah

2011-01-01

This thesis deals with electronic excitations of helium physisorbed on metal substrates. It is studied to what extent the electronic properties change compared to the gas phase due to the increased helium density and the proximity of the metal. Furthermore, the influence of different substrate materials is investigated systematically. To this end, up to two helium layers were adsorbed onto Ru (001), Pt (111), Cu (111), and Ag (111) surfaces in a custom-made cryostat. These samples were studied spectroscopically using synchrotron radiation and a time-of-flight detector. The experimental results were then analyzed in comparison with extensive theoretical model calculations.

Electronic properties of physisorbed helium

Energy Technology Data Exchange (ETDEWEB)

Kossler, Sarah

2011-09-22

This thesis deals with electronic excitations of helium physisorbed on metal substrates. It is studied to what extent the electronic properties change compared to the gas phase due to the increased helium density and the proximity of the metal. Furthermore, the influence of different substrate materials is investigated systematically. To this end, up to two helium layers were adsorbed onto Ru (001), Pt (111), Cu (111), and Ag (111) surfaces in a custom-made cryostat. These samples were studied spectroscopically using synchrotron radiation and a time-of-flight detector. The experimental results were then analyzed in comparison with extensive theoretical model calculations.

Laser spectroscopy of antiprotonic helium

CERN Document Server

Hori, M

2005-01-01

When antiprotons (i.e. the antimatter counterpart of protons) are stopped in helium gas, 97% of them annihilate within picoseconds by reacting with the helium nuclei; a 3% fraction, however, survive with an anomalously long lifetime of several microseconds. This longevity is due to the formation of antiprotonic helium, which is a three-body Rydberg atom composed of an antiproton, electron, and helium nucleus. The ASACUSA experimental collaboration has recently synthesized large numbers of these atoms using CERN's Antiproton Decelerator facility, and measured the atom's transition frequencies to 60 parts per billion by laser spectroscopy. By comparing the experimental results with recent three-body QED calculations and the known antiproton cyclotron frequency, we were able to show that the antiproton mass and charge are the same as the corresponding proton values to a precision of 10 parts per billion. Ongoing and future series of experiments will further improve the experimental precision by using chirp-compe...

Femtosecond spectroscopy on alkali-doped helium nanodroplets; Femtosekundenspektroskopie an alkalidotierten Helium-Nanotroepfchen

Energy Technology Data Exchange (ETDEWEB)

Claas, P.

2006-01-15

In the present thesis first studies on the short-time dynamics in alkali dimers and microclusters, which were bound on the surface of superfluid helium droplets, were presented. The experiments comprehended pump-probe measurements on the fs scale on the vibration dynamics on the dimers and on the fragmentation dynamics on the clusters. Generally by the studies it was shown that such extremely short slopes can also be observed on helium droplets by means of the femtosecond spectroscopy.

Helium 3 precipitation in AISI 316L stainless steel induced by radioactive decay of tritium: Microstructural study of helium bubble precipitation

International Nuclear Information System (INIS)

Brass, A.M.; Chene, J.

1994-01-01

The development of the thermonuclear technology has given rise to a renewed interest in the study of the behavior of helium in metals. A great amount of work is still required for the understanding of the role of helium on the mechanical properties of structural materials for fusion technology, especially austenitic stainless steels. This article deals with the study of the influence of thermomechanical heat treatments, aging conditions (temperature and time), and helium concentration of helium bubble precipitation in a 316L austenitic steel. Helium was generated by the radioactive decay of tritium (tritium trick). Helium bubbles impede the grain growth in 316L steel aged at 1,373 K and also the recrystallization reaction at this temperature if cold working is performed prior to aging. Transmission electron microscopy (TEM) observations indicated a weak helium precipitation at 1,073 and 1,223 K, presumably due to the presence of trapping sites for tritium, and no bubble growth after aging up to 100 hours. Precipitation sites are mainly dislocations in the matrix at 1,073 K and grain boundaries and individual dislocations in the matrix at 1,223 K. The large bubble size (50 nm) observed at 1,373 K, even for short aging times (0.083), can partly be attributed to bubble dragging by dislocations toward the grain boundaries. Cold deformation prior to aging leads to a larger bubble size due to growth enhancement during recrystallization. Decreasing the helium content leads to a smaller helium bubble size and density. Tritium trapping at helium bubbles may favor helium 3 accumulation on defects such as grain boundaries, as observed by tritium autoradiography

Beryllium irradiation embrittlement test programme. Material and specimen specification, manufacture and qualification

International Nuclear Information System (INIS)

Harries, D.R.; Dalle Donne, M.

1996-06-01

The report presents the specification, manufacture and qualification of the beryllium specimens to be irradiated in the BR2 reactor in Mol to investigate the effect of the neutron irradiation on the embrittlement as a function of temperature and beryllium oxide content. This work was been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Union within the European Fusion Technology Program. (orig.)

The adsorption of helium atoms on coronene cations

Energy Technology Data Exchange (ETDEWEB)

Kurzthaler, Thomas; Rasul, Bilal; Kuhn, Martin; Scheier, Paul, E-mail: Paul.Scheier@uibk.ac.at, E-mail: andrew.ellis@le.ac.uk [Institut für Ionenphysik und Angewandte Physik, Universität Innsbruck, Technikerstr. 25, A-6020 Innsbruck (Austria); Lindinger, Albrecht [Institut für Experimentalphysik, Freie Universität Berlin, Arnimallee 14, 14195 Berlin (Germany); Ellis, Andrew M., E-mail: Paul.Scheier@uibk.ac.at, E-mail: andrew.ellis@le.ac.uk [Department of Chemistry, University of Leicester, University Road, Leicester LE1 7RH (United Kingdom)

2016-08-14

We report the first experimental study of the attachment of multiple foreign atoms to a cationic polycyclic aromatic hydrocarbon (PAH). The chosen PAH was coronene, C{sub 24}H{sub 12}, which was added to liquid helium nanodroplets and then subjected to electron bombardment. Using mass spectrometry, coronene cations decorated with helium atoms were clearly seen and the spectrum shows peaks with anomalously high intensities (“magic number” peaks), which represent ion-helium complexes with added stability. The data suggest the formation of a rigid helium layer consisting of 38 helium atoms that completely cover both faces of the coronene ion. Additional magic numbers can be seen for the further addition of 3 and 6 helium atoms, which are thought to attach to the edge of the coronene. The observation of magic numbers for the addition of 38 and 44 helium atoms is in good agreement with a recent path integral Monte Carlo prediction for helium atoms on neutral coronene. An understanding of how atoms and molecules attach to PAH ions is important for a number of reasons including the potential role such complexes might play in the chemistry of the interstellar medium.

Diagnostic experimental results on the hydrogen embrittlement of austenitic steels

Energy Technology Data Exchange (ETDEWEB)

Gavriljuk, V.G.; Shivanyuk, V.N.; Foct, J

2003-03-14

Three main available hypotheses of hydrogen embrittlement are analysed in relation to austenitic steels based on the studies of the hydrogen effect on the interatomic bonds, phase transformations and microplastic behaviour. It is shown that hydrogen increases the concentration of free electrons, i.e. enhances the metallic character of atomic interactions, although such a decrease in the interatomic bonding cannot be a reason for brittleness and rather assists an increased plasticity. The hypothesis of the critical role of the hydrogen-induced {epsilon} martensite was tested in the experiment with the hydrogen-charged Si-containing austenitic steel. Both the fraction of the {epsilon} martensite and resistance to hydrogen embrittlement were increased due to Si alloying, which is at variance with the pseudo-hydride hypothesis. The hydrogen-caused early start of the microplastic deformation and an increased mobility of dislocations, which are usually not observed in the common mechanical tests, are revealed by the measurements of the strain-dependent internal friction, which is consistent with the hypothesis of the hydrogen-enhanced localised plasticity. An influence of alloying elements on the enthalpy E{sub H} of hydrogen migration in austenitic steels is studied using the temperature-dependent internal friction and a correlation is found between the values of E{sub H} and hydrogen-caused decrease in plasticity. A mechanism for the transition from the hydrogen-caused microplasticity to the apparent macrobrittle fracture is proposed based on the similarity of the fracture of hydrogenated austenitic steels to that of high nitrogen steels.

Embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

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

1996-01-01

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

Hydrogen embrittlement of ASTM A 203 D nuclear structural steel

International Nuclear Information System (INIS)

Chakravartty, J.K.; Prasad, G.E.; Sinha, T.K.; Asundi, M.K.

1986-01-01

The influence of hydrogen on the mechanical properties of ASTM A 203 D nuclear structural steel has been studied by tension, bend and delayed-failure tests at room temperature. While the tension tests of hydrogen charged unnotched specimens reveal no change in ultimate strength and ductility, the effect of hydrogen is manifested in notched specimens (tensile and bend) as a decrease in ultimate strength (maximum load in bend test) and ductility; the effect increases with increasing hydrogen content. It is observed that for a given hydrogen concentration, the decrease in bend ductility is remarkably large compared to that in tensile ductility. Hydrogen charging does not cause any delayed-failure upto 200 h under an applied tensile stress, 0.85 times the notch tensile strength. However delayed failure occurs in hydrogen charged bend samples in less than 10 h under an applied bending load of about 0.80 times of the uncharged maximum load. Fractographs of hydrogen charged unnotched specimens show ductile dimple fracture, while those of notched tension and bend specimens under hydrogen-charged conditions show a mixture of ductile dimple and quasi-cleavage cracking. The proportion of quasi-cleavage cracking increases with increasing hydrogen content and this fracture mode is more predominant in bend specimens. The changes in tensile properties and fracture modes can reasonably be explained by existing theories of hydrogen embrittlement. An attempt is made to explain the significant difference in the embrittlement susceptibility of bend and tensile specimens in the light of difference in triaxiality and plastic zone size near the notch tip. (orig.)

Helium behavior in oxide dispersion strengthened (ODS) steel: Insights from ab initio modeling

Science.gov (United States)

Sun, Dan; Li, Ruihuan; Ding, Jianhua; Huang, Shaosong; Zhang, Pengbo; Lu, Zheng; Zhao, Jijun

2018-02-01

Using first-principles calculations, we systemically investigate the energetics and stability behavior of helium (He) atoms and small Hen (n = 2-4) clusters inside oxide dispersion strengthened (ODS) steel, as well as the incorporation of large amount of He atoms inside Y2O3 crystal. From the energetic point of view, He atom inside Y2O3 cluster is most stable, followed by the interstitial sites at the α-Fe/Y2O3 interface, and the tetrahedral interstitial sites inside α-Fe region. We further consider Hen (n = 2-4) clusters at the tetrahedral interstitial site surrounded by four Y atoms, which is the most stable site in the ODS steel model. The incorporation energies of all these Hen clusters are lower than that of single He atom in α-Fe, while the binding energy between two He atoms is relatively small. With insertion of 15 He atoms into 80-atom unit cell of Y2O3 crystal, the incorporation energy of He atoms is still lower than that of He4 cluster in α-Fe crystal. These theoretical results suggest that He atoms tend to aggregate inside Y2O3 clusters or at the α-Fe/Y2O3 interface, which is beneficial to prevent the He embrittlement in ODS steels.

Helium release from radioisotope heat sources

Energy Technology Data Exchange (ETDEWEB)

Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

1984-05-01

Diffusion of helium in /sup 238/PuO/sub 2/ fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel.

Recent developments in liquid helium 3

International Nuclear Information System (INIS)

Ramarao, I.

1977-01-01

The current status of the theories for the ground state of liquid helium 3, are reviewed. To begin with, a brief summary of the experimental results on the thermodynamic properties of liquid helium 3 including its recently discovered superfulid phases is given. The basic ideas of the Landau theory of a normal Fermi liquid are then introduced. A qualitative discussion of the current understanding of the anisotropic phases of superfluid helium 3 is given, the microscopic calculaations for the binding energy of liquid helium 3 are reviewed and the results obtained for the two-body contributions to the binding energy using the Brueckner-Goldstone formulation and that of Mohling and his collaborators are summarized and discussed. The importance of a proper estimate of the three-body contributions to the binding energy is stressed. The results obtained in the literature using variational methods and constrained variational methods are discussed. A critical analysis of the results by various methods is given. Despite much effort the basic problem of the ground state of liquid helium 3, remains unresolved. (author)

Helium release from radioisotope heat sources

International Nuclear Information System (INIS)

Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

1984-05-01

Diffusion of helium in 238 PuO 2 fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel

Cosmological helium production simplified

International Nuclear Information System (INIS)

Bernstein, J.; Brown, L.S.; Feinberg, G.

1988-01-01

We present a simplified model of helium synthesis in the early universe. The purpose of the model is to explain clearly the physical ideas relevant to the cosmological helium synthesis, in a manner that does not overlay these ideas with complex computer calculations. The model closely follows the standard calculation, except that it neglects the small effect of Fermi-Dirac statistics for the leptons. We also neglect the temperature difference between photons and neutrinos during the period in which neutrons and protons interconvert. These approximations allow us to express the neutron-proton conversion rates in a closed form, which agrees to 10% accuracy or better with the exact rates. Using these analytic expressions for the rates, we reduce the calculation of the neutron-proton ratio as a function of temperature to a simple numerical integral. We also estimate the effect of neutron decay on the helium abundance. Our result for this quantity agrees well with precise computer calculations. We use our semi-analytic formulas to determine how the predicted helium abundance varies with such parameters as the neutron life-time, the baryon to photon ratio, the number of neutrino species, and a possible electron-neutrino chemical potential. 19 refs., 1 fig., 1 tab

Low-temperature centrifugal helium compressor

International Nuclear Information System (INIS)

Kawada, M.; Togo, S.; Akiyama, Y.; Wada, R.

1974-01-01

A centrifugal helium compressor with gas bearings, which can be operated at the temperature of liquid nitrogen, has been investigated. This compressor has the advantages that the compression ratio should be higher than the room temperature operation and that the contamination of helium could be eliminated. The outer diameter of the rotor is 112 mm. The experimental result for helium gas at low temperature shows a flow rate of 47 g/s and a compression ratio of 1.2 when the inlet pressure was 1 ata and the rotational speed 550 rev/s. The investigation is now focused on obtaining a compression ratio of 1.5. (author)

Retention of hydrogen isotopes and helium in nickel

Energy Technology Data Exchange (ETDEWEB)

Okada, Mitsumasa; Sato, Rikiya; Yamaguchi, Kenji; Yamawaki, Michio [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

1996-10-01

In the present study, a thin foil of nickel was irradiated by H{sub 2}{sup +}, D{sub 2}{sup +} and He{sup +} to a fluence of 1.2-6.0x10{sup 20}/m{sup 2} using the TBTS (Tritium Beam Test System) apparatus. The thermal desorption spectroscopy (TDS) technique was employed to evaluate the total amount of retained hydrogen isotope and helium atoms in nickel. In the spectra, two peaks appeared at 440-585K and 720-735K for helium. Hydrogen isotopes irradiation after helium preirradiation were found to enhance the helium release and to decrease the peak temperatures. Helium irradiation after hydrogen isotopes preirradiation were found to enhance the helium release, but the peak temperature showed little difference from that without preirradiation. (author)

Liquid metal embrittlement. From basic concepts to recent results related to structural materials for liquid metal spallation targets

International Nuclear Information System (INIS)

Gorse, D.; Goryachev, S.; Auger, T.

2003-01-01

At first, the basic features of LME are recalled (definition, characteristics, embrittling couples), together with classical experimental features and open questions. Then, a review of a few very recent results obtained on classical embrittling couples but using new powerful investigation techniques developed in France is proposed. Second we define LMC. The 'LME-LMC' correlation is postulated. Then we concentrate on the LME-LMC problem related to the build-up of the Liquid Metal Spallation target in the frame of the MEGAPIE project. The Russian expertise on LME is briefly mentioned. Then we present some results obtained in the frame of the Groupement de Recherche' GEDEON, focusing on steel grade T91 in contact with lead and lead-bismuth eutectic, in agreement with Russian literature. (author)

Localization of electromagnetic field on the “Brouwer-island” and liquid metal embrittlement

Energy Technology Data Exchange (ETDEWEB)

Maksimenko, V.V.; Zagaynov, V.A. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 31, Kashirskoe shosse, 115409 Moscow (Russian Federation); Karpov Institute of Physical Chemistry, Vorontsovo Pole, 10, 105064 Moscow (Russian Federation); Agranovski, I.E., E-mail: I.Agranovski@griffith.edu.au [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 31, Kashirskoe shosse, 115409 Moscow (Russian Federation); School of Engineering, Griffith University, Brisbane, 4111 QLD (Australia)

2015-03-01

Liquid metal embrittlement (LME) manifests itself as a sudden destruction of a metal sample if it is covered by a thin liquid film of eutectic mixture of specially selected metals. The proposed theoretical model of this phenomenon is based on an assumption related to the possibility of electromagnetic field localization in folds of interface between the phases or components of eutectic mixture filling cracks in solid metal surface (the typical example is In–Ga eutectic on Al-surface). Based on simultaneous presence of three different components in each space point of eutectic mixture (homogeneous In + Ga melt, solid In, and solid Ga), the system of interface folds could be simulated by the Brouwer surface – well known in topology. This surface separates three different components presented at each of its point. Such fractal surfaces posses by a finite volume. The volume occupied by the surface is defined as a difference between the eutectic mixture volume and the sum of volumes of its components. We investigate localization of external electromagnetic radiation in this system of folds. Due to very large magnitude of effective dielectric permeability of the considered system, at relative small volume change and fractal dimension of interface close to the value 3, the wave length of incident radiation inside the system is considerably decreased and multiscale folds are filled with localized photons. A probability of this process and the life time of the localized photons are calculated. The localized photons play crucial role in destruction of primary cracks in the metal surface. They are capable “to switch of” the Coulomb attraction of charge fluctuations on opposite “banks” of the crack filled with the eutectic. As a result, the crack could break down. - Highlights: • A new theoretical model of liquid metal embrittlement has been developed. • Light localization has a strong influence on liquid metal embrittlement. • Light is localized in folds at

Localization of electromagnetic field on the “Brouwer-island” and liquid metal embrittlement

International Nuclear Information System (INIS)

Maksimenko, V.V.; Zagaynov, V.A.; Agranovski, I.E.

2015-01-01

Liquid metal embrittlement (LME) manifests itself as a sudden destruction of a metal sample if it is covered by a thin liquid film of eutectic mixture of specially selected metals. The proposed theoretical model of this phenomenon is based on an assumption related to the possibility of electromagnetic field localization in folds of interface between the phases or components of eutectic mixture filling cracks in solid metal surface (the typical example is In–Ga eutectic on Al-surface). Based on simultaneous presence of three different components in each space point of eutectic mixture (homogeneous In + Ga melt, solid In, and solid Ga), the system of interface folds could be simulated by the Brouwer surface – well known in topology. This surface separates three different components presented at each of its point. Such fractal surfaces posses by a finite volume. The volume occupied by the surface is defined as a difference between the eutectic mixture volume and the sum of volumes of its components. We investigate localization of external electromagnetic radiation in this system of folds. Due to very large magnitude of effective dielectric permeability of the considered system, at relative small volume change and fractal dimension of interface close to the value 3, the wave length of incident radiation inside the system is considerably decreased and multiscale folds are filled with localized photons. A probability of this process and the life time of the localized photons are calculated. The localized photons play crucial role in destruction of primary cracks in the metal surface. They are capable “to switch of” the Coulomb attraction of charge fluctuations on opposite “banks” of the crack filled with the eutectic. As a result, the crack could break down. - Highlights: • A new theoretical model of liquid metal embrittlement has been developed. • Light localization has a strong influence on liquid metal embrittlement. • Light is localized in folds at

Helium sources to groundwater in active volcanic terrain, and implications for tritium-helium dating at Mount St. Helens

Energy Technology Data Exchange (ETDEWEB)

Gates, John B. [Department of Earth and Atmospheric Sciences, University of Nebraska-Lincoln, 217 Bessey Hall, Lincoln NE 68588 (United States)

2013-07-01

Groundwater helium sources and residence times were investigated using groundwater discharging from springs surrounding Mount St. Helens in the Cascades region of the United States. Significant contributions of mantle helium were found in all samples and are attributable to interaction between groundwater and magmatic gases. Bounding calculations for residence times were made on the basis of helium isotope mixing plots and historical tritium data. (authors)

Explosive helium burning in white dwarf stars

Energy Technology Data Exchange (ETDEWEB)

Khokhlov, A.M. (AN SSSR, Moscow. Astronomicheskij Sovet)

1984-04-01

Helium burning kinetics in white dwarfs has been considered at constant temperatures T >= 10/sup 9/ K and densities rho >10/sup 5/ g/cm/sup 3/. It is found, that helium detonation in white dwarfs does not lead to formation of light (A < 56) elements. Thus, helium white dwarf model for supernova 1 is inconsistent with observations.

A new helium gas recovery and purification system

International Nuclear Information System (INIS)

Yamamotot, T.; Suzuki, H.; Ishii, J.; Hamana, I.; Hayashi, S.; Mizutani, S.; Sanjo, S.

1974-01-01

A helium gas recovery and purification system, based on the principle of gas permeation through a membrane, is described. The system can be used for the purification of helium gas containing air as a contaminant. The apparatus, operating at ambient temperature does not need constant attention, the recovery ratio of helium gas is satisfactory and running costs are low. Gases other than helium can be processed with the apparatus. (U.K.)

Investigation of the delay in pressure vessel embrittlement specimen analysis for the Oak Ridge National Laboratory High Flux Isotope Reactor

International Nuclear Information System (INIS)

Rothrock, J.D.; Hoffman, E.E.; Manthey, G.C.; Sheffey, D.W.

1987-01-01

Analysis of the investigative data pertaining to this incident reveals the following conditions as key findings and probable causes: (1) The contractor failed to properly implement the surveillance program for monitoring reactor pressure vessel embrittlement. (2) Contractor and DOE organizations provided less than adequate oversight and independent overview, especially by not requiring operating organizations to provide documented evidence to substantiate claims that there was ''no problem'' with respect to embrittlement. (3) Although the temperature limitation for reactor pressurization identified in the Technical Specifications was never violated, the basis of this safety limitation was violated. (4) The basis for concluding that there would be no embrittlement of the pressure vessel steel over the expected life of the reactor is questionable. (5) The contractor and DOE failed to make the surveillance program visible by incorporating it in the Technical Specifications. (6) The Accident Analysis/Final Safety Analysis Report was never adequately reviewed and updated subsequent to its initial issuance. (7) Surveillance specimen analysis was incomplete and never transmitted to reactor operating personnel in a usable format prior to November 1986. (8) There was extensive delays (many years) in the testing, analysis, and reporting of surveillance program results

Helium leak and chemical impurities control technology in HTTR

International Nuclear Information System (INIS)

Tochio, Daisuke; Shimizu, Atsushi; Hamamoto, Shimpei; Sakaba, Nariaki

2014-01-01

Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly. The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system. This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown. (author)

Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

Science.gov (United States)

Larson, V. R.; Gunn, S. V.; Lee, J. C.

1975-01-01

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

A reciprocating liquid helium pump used for forced flow of supercritical helium

International Nuclear Information System (INIS)

Krafft, G.; Zahn, G.

1978-01-01

The performance of a small double acting piston pump for circulating helium in a closed heat transfer loop is described. The pump was manufactured by LINDE AG, Munich, West Germany. The measured flow rate of supercritical helium was about 17 gs -1 (500 lhr -1 ) with a differential pressure of Δp = 0.5 x 10 5 Nm -2 at a working pressure of p = 6 x 10 5 Nm -2 . At differential pressures beyond 0.5 x 10 5 Nm -2 the volumetric efficiency decreases. (author)

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

International Nuclear Information System (INIS)

Brooks, J.N.; Krauss, A.; Mattas, R.F.; Smith, D.L.; Nygren, R.E.; Doyle, B.L.; McGrath, R.T.; Walsh, D.; Dippel, K.H.; Finken, K.H.

1990-01-01

A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, ≅ 25x25 cm 2 heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of ≅ 50A nickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a ≅ 3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions. In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has been measured. The results appear favorable, showing high helium trapping (≅ 10-50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991. (orig.)

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

International Nuclear Information System (INIS)

Brooks, J.N.; Krauss, A.; Mattas, R.F.; Smith, D.L.; Nygren, R.E.; Doyle, B.L.; McGrath, R.T.; Walsh, D.; Dippel, K.H.; Finken, K.H.

1990-01-01

A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, ∼25 x 25 cm 2 heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of ∼50 angstrom nickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a ∼3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions. In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has bee measured. The results appear favorable, showing high helium trapping (∼10--50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991. 12 refs., 2 figs., 2 tabs

Sonic Helium Detectors in the Fermilab Tevatron

Science.gov (United States)

Bossert, R. J.

2006-04-01

In the Fermilab Tevatron cryogenic system there are many remotely located low-pressure plate relief valves that must vent large volumes of cold helium gas when magnet quenches occur. These valves can occasionally stick open or not reseat completely, resulting in a large helium loss. As such, the need exists for a detector to monitor the relief valve's discharge area for the presence of helium. Due to the quantity needed, cost is an important factor. A unit has been developed and built for this purpose that is quite inexpensive. Its operating principle is based on the speed of sound, where two closely matched tubes operate at their acoustic resonant frequency. When helium is introduced into one of these tubes, the resulting difference in acoustic time of flight is used to trigger an alarm. At present, there are 39 of these units installed and operating in the Tevatron. They have detected many minor and major helium leaks, and have also been found useful in detecting a rise in the helium background in the enclosed refrigerator buildings. This paper covers the construction, usage and operational experience gained with these units over the last several years.

Sonic helium detectors in the Fermilab Tevatron

International Nuclear Information System (INIS)

Bossert, R.J.; Fermilab

2006-01-01

In the Fermilab Tevatron cryogenic system there are many remotely located low-pressure plate relief valves that must vent large volumes of cold helium gas when magnet quenches occur. These valves can occasionally stick open or not reseat completely, resulting in a large helium loss. As such, the need exists for a detector to monitor the relief valve's discharge area for the presence of helium. Due to the quantity needed, cost is an important factor. A unit has been developed and built for this purpose that is quite inexpensive. Its operating principle is based on the speed of sound, where two closely matched tubes operate at their acoustic resonant frequency. When helium is introduced into one of these tubes, the resulting difference in acoustic time of flight is used to trigger an alarm. At present, there are 39 of these units installed and operating in the Tevatron. They have detected many minor and major helium leaks, and have also been found useful in detecting a rise in the helium background in the enclosed refrigerator buildings. This paper covers the construction, usage and operational experience gained with these units over the last several years

Helium production in reactor materials

International Nuclear Information System (INIS)

Lippincott, E.P.; McElroy, W.N.; Farrar, H. IV.

1975-02-01

Comparisons of integral helium production measurements with predictions based on ENDF/B Version IV cross sections have been made. It is concluded that an ENDF/B helium production cross section file should be established in order to ensure a complete and consistent cross section evaluation to meet accuracies required for LMFBR, CTR, and LWR applications. (U.S.)

Pressurized helium II-cooled magnet test facility

International Nuclear Information System (INIS)

Warren, R.P.; Lambertson, G.R.; Gilbert, W.S.; Meuser, R.B.; Caspi, S.; Schafer, R.V.

1980-06-01

A facility for testing superconducting magnets in a pressurized bath of helium II has been constructed and operated. The cryostat accepts magnets up to 0.32 m diameter and 1.32 m length with current to 3000 A. In initial tests, the volume of helium II surrounding the superconducting magnet was 90 liters. Minimum temperature reached was 1.7 K at which point the pumping system was throttled to maintain steady temperature. Helium II reservoir temperatures were easily controlled as long as the temperature upstream of the JT valve remained above T lambda; at lower temperatures control became difficult. Positive control of the temperature difference between the liquid and cold sink by means of an internal heat source appears necessary to avoid this problem. The epoxy-sealed vessel closures, with which we have had considerable experience with normal helium vacuum, also worked well in the helium II/vacuum environment

Modeling Secondary Neutral Helium in the Heliosphere

International Nuclear Information System (INIS)

Müller, Hans-Reinhard; Möbius, Eberhard; Wood, Brian E.

2016-01-01

An accurate, analytic heliospheric neutral test-particle code for helium atoms from the interstellar medium (ISM) is coupled to global heliospheric models dominated by hydrogen and protons from the solar wind and the ISM. This coupling enables the forward-calculation of secondary helium neutrals from first principles. Secondaries are produced predominantly in the outer heliosheath, upwind of the heliopause, by charge exchange of helium ions with neutral atoms. The forward model integrates the secondary production terms along neutral trajectories and calculates the combined neutral helium phase space density in the innermost heliosphere where it can be related to in-situ observations. The phase space density of the secondary component is lower than that of primary neutral helium, but its presence can change the analysis of primaries and the ISM, and can yield valuable insight into the characteristics of the plasma in the outer heliosheath. (paper)

Behaviour of helium after implantation in molybdenum

International Nuclear Information System (INIS)

Viaud, C.; Maillard, S.; Carlot, G.; Valot, C.; Gilabert, E.; Sauvage, T.; Peaucelle, C.; Moncoffre, N.

2009-01-01

This study deals with the behaviour of helium in a molybdenum liner dedicated to the retention of fission products. More precisely this work contributes to evaluate the release of implanted helium when the gas has precipitated into nanometric bubbles close to the free surface. A simple model dedicated to calculate the helium release in such a condition is presented. The specificity of this model lays on the assumption that the gas is in equilibrium with a simple distribution of growing bubbles. This effort is encouraging since the calculated helium release fits an experimental dataset with a set of parameters in good agreement with the literature

Use of helium in uranium exploration, Grants district

International Nuclear Information System (INIS)

DeVoto, R.H.; Mead, R.H.; Martin, J.P.; Bergquist, L.E.

1980-01-01

The continuous generation of inert helium gas from uranium and its daughter products provides a potentially useful means for remote detection of uranium deposits. The practicality of conducting helium surveys in the atmosphere, soil gas, and ground water to explore for buried uranium deposits has been tested in the Grants district and in the Powder River Basin of Wyoming. No detectable helium anomalies related to buried or surface uranium deposits were found in the atmosphere. However, reproducible helium-in-soil-gas anomalies were detected spatially related to uranium deposits buried from 50 to 800 ft deep. Diurnal and atmospheric effects can cause helium content variations (noise) in soil gas that are as great as the anomalies observed from instantaneous soil-gas samples. Cumulative soil-gas helium analyses, such as those obtained from collecting undisturbed soil samples and degassing them in the laboratory, may reveal anomalies from 5 to 100 percent above background. Ground water samples from the Grants district, New Mexico, and the Powder River Basin, Wyoming, have distinctly anomalous helium values spatially related to buried uranium deposits. In the southern Powder River Basin, helium values 20 to 200 percent above background occur 2 to 18 mile down the ground-water flow path from known uranium roll-front deposits. In the Grants district, helium contents 40 to 700 percent above background levels are present in ground waters from the host sandstone in the vicinity of uranium deposits and from aquifers up to 3,000 ft stratigraphically above the deep uranium deposits. The use of helium in soil and ground-water surveys, along with uranium and radon analyses of the same materials, is strongly recommended is expensive, deep, uranium-exploration programs such as those being conducted in the Grants district

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

Energy Technology Data Exchange (ETDEWEB)

Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation); National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, Moscow 115409 (Russian Federation); Gurovich, B.A.; Bukina, Z.V.; Frolov, A.S.; Maltsev, D.A.; Krikun, E.V.; Zhurko, D.A.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

2017-07-15

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

Manufacturing cycle for pure neon-helium mixture production

International Nuclear Information System (INIS)

Batrakov, B.P.; Kravchenko, V.A.

1980-01-01

The manufacturing cycle for pure neon-helium mixture production with JA-300 nitrogen air distributing device has been developed. Gas mixture containing 2-3% of neon-helium mixture (the rest is mainly nitrogen 96-97%) is selected out of the cover of the JA-300 column condensator and enters the deflegmator under the 2.3-2.5 atm. pressure. The diflegmator presents a heat exchange apparatus in which at 78 K liquid nitrogen the condensation of nitrogen from the mixture of gases entering from the JA-300 column takes place. The enriched gas mixture containing 65-70% of neon-helium mixture and 30-35% of nitrogen goes out from the deflegmator. This enriched neon-helium mixture enters the gasgoeder for impure (65-70%) neon-helium mixture. Full cleaning of-neon helium mixture of nitrogen is performed by means of an adsorber. As adsorbent an activated coal has been used. Adsorption occurs at the 78 K temperature of liquid nitrogen and pressure P=0.1 atm. As activated coal cooled down to nitrogen temperature adsorbs nitrogen better than neon and helium, the nitrogen from the mixture is completely adsorbed. Pure neon-helium mixture from the adsorber comes into a separate gasgolder. In one campaign the cycle allows obtaining 2 m 3 of the mixture. The mixture contains 0.14% of nitrogen, 0.01% of oxygen and 0.06% of hydrogen

Standard Guide for Simulation of Helium Effects in Irradiated Metals

CERN Document Server

American Society for Testing and Materials. Philadelphia

1996-01-01

1.1 This guide provides advice for conducting experiments to investigate the effects of helium on the properties of metals where the technique for introducing the helium differs in some way from the actual mechanism of introduction of helium in service. Simulation techniques considered for introducing helium shall include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended. 1.2 Two other methods for introducing helium into irradiated materials are not covered in this guide. They are the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, and isotopic tailoring in both fast and mixed-spectrum fission reactors. These techniques are described in Refs (1-5). Dual ion beam techniques (6) for simultaneously implanting helium and generating displacement damage are also not included here. This lat...

Helium-induced weld degradation of HT-9 steel

International Nuclear Information System (INIS)

Wang, Chin-An; Chin, B.A.; Lin, Hua T.; Grossbeck, M.L.

1992-01-01

Helium-bearing Sandvik HT-9 ferritic steel was tested for weldability to simulate the welding of structural components of a fusion reactor after irradiation. Helium was introduced into HT-9 steel to 0.3 and 1 atomic parts per million (appm) by tritium doping and decay. Autogenous single pass full penetration welds were produced using the gas tungsten arc (GTA) welding process under laterally constrained conditions. Macroscopic examination showed no sign of any weld defect in HT-9 steel containing 0.3 appm helium. However, intergranular micro cracks were observed in the HAZ of HT-9 steel containing 1 appm helium. The microcracking was attributed to helium bubble growth at grain boundaries under the influence of high stresses and temperatures that were present during welding. Mechanical test results showed that both yield strength (YS) and ultimate tensile strength (UTS) decreased with increasing temperature, while the total elongation increased with increasing temperature for all control and helium-bearing HT-9 steels

Gaseous hydrogen embrittlement of an API X80 ferrito-pearlitic steel; Fragilisation par l'hydrogene gazeux d'un acier ferrito-perlitique de grade API X80

Energy Technology Data Exchange (ETDEWEB)

Moro, I.

2009-11-15

This work deals with hydrogen embrittlement, at ambient temperature and under a high pressure gaseous way, of an API X80 high elasticity limit steel used for pipelines construction, and with the understanding of the associated physical mechanisms of the embrittlement. At first has been described a bibliographic study of the adsorption, absorption, diffusion, transport and trapping of hydrogen in the steels. Then has been carried out an experimental and numerical study concerning the implantation in the finite element code CASTEM3M of a hydrogen diffusion model coupled to mechanical fields. The hydrogen influence on the mechanical characteristics of the X80 steel, of a ferrito-pearlitic microstructure has been studied with tensile tests under 300 bar of hydrogen and at ambient temperature. The sensitivity of the X80 steel to hydrogen embrittlement has been analyzed by tensile tests at different deformation velocities and under different hydrogen pressures on axisymmetrical notched test specimens. These studies show that the effect of the hydrogen embrittlement vary effectively with the experimental conditions. Moreover, correlated with the results of the tests simulations, it has been shown too that in these experimental conditions and for that steel, the hydrogen embrittlement is induced by three different hydrogen populations: the hydrogen trapped at the ferrite/perlite interfaces, the hydrogen adsorbed on surface and the reticular hydrogen trapped in the material volume. At last, the tensile and rupture tests of specimens, during which atmosphere changes have been carried out, have shown a strong reversibility of the hydrogen embrittlement, associated with its initiation as soon as hydrogen is introduced in the atmosphere. At last, three hydrogen mechanisms, depending of the different hydrogen populations are presented and discussed. (O.M.)

Effect of Low-Temperature Sensitization on Hydrogen Embrittlement of 301 Stainless Steel

OpenAIRE

Chieh Yu; Ren-Kae Shiue; Chun Chen; Leu-Wen Tsay

2017-01-01

The effect of metastable austenite on the hydrogen embrittlement (HE) of cold-rolled (30% reduction in thickness) 301 stainless steel (SS) was investigated. Cold-rolled (CR) specimens were hydrogen-charged in an autoclave at 300 or 450 °C under a pressure of 10 MPa for 160 h before tensile tests. Both ordinary and notched tensile tests were performed in air to measure the tensile properties of the non-charged and charged specimens. The results indicated that cold rolling caused the transforma...

Influence of pre-hydriding on embrittlement of E110 alloy under LOCA conditions

International Nuclear Information System (INIS)

VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Fedotov, P.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Kuznetsov, V.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Nechaeva, O.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Novikov, V.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Salatov, A.; Ignatiev, D.; Mokrushin, A.; Soldatkin, D.; Urusov, A.

2015-01-01

The researches presented in this paper were carried out in the framework of TVS-K project developed by JSC “TVEL”. The data on the corrosion and residual ductility of unirradiated and pre-hydrided E110 alloy under LACA conditions at temperature range from 1100 to 1200°C are presented. The hydrogen concentration was varied from 30 (as-received) to 600 wppm. The initial concentration of hydrogen has no effect on the oxidation kinetics, while the oxidation kinetics are parabolic and the breakaway oxidation is not observed. Oxide films on surfaces of claddings are black and shining. There are no cracks, visual spots and peelings. The residual ductility of oxidised samples decrease with hydrogen concentration rise. The residual ductility of claddings oxidized at 1100 °C, generally higher than the same of the claddings oxidized at 1200 °C. E110 alloy has a good residual ductility in comparison to Zry-4, ZIRLO, M5. Joint analysis of the test results allowed us to formulate embrittlement criteria of the E110 alloy under LOCA conditions. This embrittlement criterion is preliminary, because the experimental data base must to be enlarged by results of tests with claddings of another geometry and quench experiments. (author)

Parameters promoting liquid metal embrittlement of the T91 steel in lead-bismuth eutectic alloy

International Nuclear Information System (INIS)

Proriol Serre, I.; Ye, C.; Vogt, J.B.

2015-01-01

The use of liquid lead-bismuth eutectic (LBE) as a spallation target and a coolant in accelerator-driven systems raises the question of the reliability of structural materials, such as T91 martensitic steel in terms of liquid metal assisted damage and corrosion. In this study, the mechanical behaviour of the T91 martensitic steel was examined in liquid lead-bismuth eutectic (LBE) and in inert atmosphere. Several conditions showed the most sensitive embrittlement factor. The Small Punch Test technique was employed using smooth specimens. In this standard heat treatment, T91 appeared in general as a ductile material, and became brittle in the considered conditions if the test was performed in LBE. It turns out that the loading rate appeared as a critical parameter for the occurrence of liquid metal embrittlement (LME) of the T91 steel in LBE. Loading the T91 very slowly instead of rapidly in oxygen saturated LBE resulted in brittle fracture. Furthermore, low-oxygen content in LBE and an increase in temperature promote LME. (authors)

Standard Test Method for Mechanical Hydrogen Embrittlement Evaluation of Plating/Coating Processes and Service Environments

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This test method describes mechanical test methods and defines acceptance criteria for coating and plating processes that can cause hydrogen embrittlement in steels. Subsequent exposure to chemicals encountered in service environments, such as fluids, cleaning treatments or maintenance chemicals that come in contact with the plated/coated or bare surface of the steel, can also be evaluated. 1.2 This test method is not intended to measure the relative susceptibility of different steels. The relative susceptibility of different materials to hydrogen embrittlement may be determined in accordance with Test Method F1459 and Test Method F1624. 1.3 This test method specifies the use of air melted AISI E4340 steel per SAE AMS-S-5000 (formerly MIL-S-5000) heat treated to 260 – 280 ksi (pounds per square inch x 1000) as the baseline. This combination of alloy and heat treat level has been used for many years and a large database has been accumulated in the aerospace industry on its specific response to exposure...

Effects of alloys elements, impurities and microstructural factors in austenitic stainless steel to utilize in fuel rod of nuclear reactors

International Nuclear Information System (INIS)

Yoshimoto, A.

1988-08-01

Austenitic Stainless Steel is used as cladding material of pressurized water reactor fuel rods because of its good performance. The addition of alloy elements and the control of impurities make this to happen. Fission products do not contribute to corrosion. Dimensional changes are not critical up to 1,0 x 10 22 n/cm 2 (E>0,1 MeV) of neutronic doses. The hydrogen does not cause embrittlement in the reactor operation temperatures, and helium contributes to embrittlement if the material is warmed upon 650 0 C. (author) [pt

Charged condensate and helium dwarf stars

Energy Technology Data Exchange (ETDEWEB)

Gabadadze, Gregory; Rosen, Rachel A, E-mail: gg32@nyu.edu, E-mail: rar339@nyu.edu [Center for Cosmology and Particle Physics, Department of Physics, New York University, New York, NY 10003 (United States)

2008-10-15

White dwarf stars composed of carbon, oxygen and heavier elements are expected to crystallize as they cool down below certain temperatures. Yet, simple arguments suggest that the helium white dwarf cores may not solidify, mostly because of zero-point oscillations of the helium ions that would dissolve the crystalline structure. We argue that the interior of the helium dwarfs may instead form a macroscopic quantum state in which the charged helium-4 nuclei are in a Bose-Einstein condensate, while the relativistic electrons form a neutralizing degenerate Fermi liquid. We discuss the electric charge screening, and the spectrum of this substance, showing that the bosonic long-wavelength fluctuations exhibit a mass gap. Hence, there is a suppression at low temperatures of the boson contribution to the specific heat-the latter being dominated by the specific heat of the electrons near the Fermi surface. This state of matter may have observational signatures.

On the helium gas leak test

International Nuclear Information System (INIS)

Nishikawa, Akira; Ozaki, Susumu

1975-01-01

The helium gas leak test (Helium mass spectrometer testing) has a leak detection capacity of the highest level in practical leak tests and is going to be widely applied to high pressure vessels, atomic and vacuum equipments that require high tightness. To establish a standard test procedure several series of experiments were conducted and the results were investigated. The conclusions are summarized as follows: (1) The hood method is quantitatively the most reliable method. The leak rate obtained by tests using 100% helium concentration should be the basis of the other method of test. (2) The integrating method, bell jar method, and vacuum spray method can be considered quantitative when particular conditions are satisfied. (3) The sniffer method is not to be considered quantitive. (4) The leak rate of the hood, integrating, and bell jar methods is approximately proportional to the square of the helium partial pressure. (auth.)

Hydrogen embrittlement of titanium and its alloys - a literature review

International Nuclear Information System (INIS)

Aho-Mantila, I.; Haemaelaeinen, H.

1986-05-01

Hydrogen embrittlement data of titanium and its alloys is reviewed. Especially the results obtained in spent nuclear fuel repository conditions with commercially pure titanium and TiCode-12 alloy are examined. The results show that the mechanical properties of titanium are not much affected by hydrogen when tested by smooth specimens. Much greater effects can be expected with notched fracture mechanics specimens. However, only limeted data is available. Hydrogen distribution in titanium is affected by stress, alloy composition and temperature gradients. In order to model the hydrogen-induced crack growth in titanium much more mechanistic work is needed especially to understand the behaviour of hydrogen in crack tip stress field. (author)

Ductile austenitic steel for fuel cans and core components of sodium cooled reactors; Ein duktiler austenitischer Stahl fuer Huellrohre und Kernkomponenten natriumgekuehlter Brueter

Energy Technology Data Exchange (ETDEWEB)

Schaefer, L.

1995-08-01

Two austenitic steel melts of a new composition have been studied after irradiation in the PFR fast neutron flux, in the BR2 reactor, and in the Harwell V.E. Cyclotron. The investigations were focussed on helium embrittlement and irradiation induced swelling. (orig.)

Ductile austenitic steel for fuel cans and core components of sodium cooled reactors

International Nuclear Information System (INIS)

Schaefer, L.

1995-01-01

Two austenitic steel melts of a new composition have been studied after irradiation in the PFR fast neutron flux, in the BR2 reactor, and in the Harwell V.E. Cyclotron. The investigations were focussed on helium embrittlement and irradiation induced swelling. (orig.)

Helium-3 and helium-4 acceleration by high power laser pulses for hadron therapy

Directory of Open Access Journals (Sweden)

S. S. Bulanov

2015-06-01

Full Text Available The laser driven acceleration of ions is considered a promising candidate for an ion source for hadron therapy of oncological diseases. Though proton and carbon ion sources are conventionally used for therapy, other light ions can also be utilized. Whereas carbon ions require 400 MeV per nucleon to reach the same penetration depth as 250 MeV protons, helium ions require only 250 MeV per nucleon, which is the lowest energy per nucleon among the light ions (heavier than protons. This fact along with the larger biological damage to cancer cells achieved by helium ions, than that by protons, makes this species an interesting candidate for the laser driven ion source. Two mechanisms (magnetic vortex acceleration and hole-boring radiation pressure acceleration of PW-class laser driven ion acceleration from liquid and gaseous helium targets are studied with the goal of producing 250 MeV per nucleon helium ion beams that meet the hadron therapy requirements. We show that He^{3} ions, having almost the same penetration depth as He^{4} with the same energy per nucleon, require less laser power to be accelerated to the required energy for the hadron therapy.

Superfluid helium at subcritical active core

International Nuclear Information System (INIS)

Vasil'ev, V.V.; Lopatkin, A.V.; Muratov, V.G.; Rakhno, I.L.

2002-01-01

Power range and neutron flux wherein super thermal source was realized at high volume of superfluid helium were investigated. MCU, BRAND, MCNP codes were used for the calculation of reactors. It is shown that the availability of full-size diameter for cryogenic source of ultracold neutrons, as the source with superfluid helium is considered, is possible in the reflector of subcritical assembly. Results obtained from the MCNP-4B code application demonstrated that the density of thermal neutron flux in helium must be not higher than 2.3 x 10 11 s -1 cm -2 [ru

Helium leak testing methods in nuclear applications

International Nuclear Information System (INIS)

Ahmad, Anis

2004-01-01

Helium mass-spectrometer leak test is the most sensitive leak test method. It gives very reliable and sensitive test results. In last few years application of helium leak testing has gained more importance due to increased public awareness of safety and environment pollution caused by number of growing chemical and other such industries. Helium leak testing is carried out and specified in most of the critical area applications like nuclear, space, chemical and petrochemical industries

Pressurized-helium breakdown at very low temperatures

Energy Technology Data Exchange (ETDEWEB)

Metas, R J

1972-06-01

An investigation of the electrical-breakdown behavior of helium at very low temperatures has been carried out to assist the design and development of superconducting power cables. At very high densities, both liquid and gaseous helium showed an enhancement in electric strength when pressurized to a few atmospheres; conditioned values of breakdown fields then varied between 30 and 45 MV/m. Breakdown processes occurring over a wide range of helium densities are discussed. 24 references.

Dipole moments of molecules solvated in helium nanodroplets

International Nuclear Information System (INIS)

Stiles, Paul L.; Nauta, Klaas; Miller, Roger E.

2003-01-01

Stark spectra are reported for hydrogen cyanide and cyanoacetylene solvated in helium nanodroplets. The goal of this study is to understand the influence of the helium solvent on measurements of the permanent electric dipole moment of a molecule. We find that the dipole moments of the helium solvated molecules, calculated assuming the electric field is the same as in vacuum, are slightly smaller than the well-known gas-phase dipole moments of HCN and HCCCN. A simple elliptical cavity model quantitatively accounts for this difference, which arises from the dipole-induced polarization of the helium

Helium atoms and molecules in strong magnetic fields

Science.gov (United States)

Mori, K.

Recent theoretical studies have shown that the neutron star surface may be composed of helium or heavier elements as hydrogen may be quickly depleted by diffuse nuclear burning Chang Bildsten However while Hydrogen atmospheres have been studied in great details atomic data for helium is available only for He ion Pavlov Bezchastnov 2005 We performed Hartree-Fock type calculation for Helium atom and molecules and computed their binding ionization and dissociation energies in strong magnetic fields B sim10 12 -- 10 15 G We will present ionization balance of Helium atmospheres at typical magnetic field strengths and temperatures to radio-quiet neutron stars and AXPs We will also discuss several implications of helium atmosphere to X-ray data of isolated neutron stars focusing on the detected spectral features

Tritium and helium-3 in metals

International Nuclear Information System (INIS)

Lasser, R.

1989-01-01

The book surveys recent results on the behaviour of tritium and its decay product helium-3 metals. In contrast to many earlier books which discuss the properties of the stable hydrogen isotopes without mentioning tritium, this book reviews mainly the results on tritium in metals. Due to the difficulties in preparing metal tritide samples, very important quantities such as diffusivity, superconductivity, solubility, etc. have only been determined very recently. The book not only presents the measured tritium data, but also the isotopic dependency of the different physical properties by comparing H, D and T results. A chapter is devoted to helium-3 in metals. Aspects such as helium release, generation of helium bubbles, swelling, and change of the lattice parameter upon aging are discussed. The book provides the reader with up-to-date information and deep insight into the behaviour of H, D, T and He-3 in metals. Further important topics such a tritium production, its risks, handling and discharge to the environment are also addressed

Helium cosmic ray flux measurements at Mars

International Nuclear Information System (INIS)

Lee, Kerry; Pinsky, Lawrence; Andersen, Vic; Zeitlin, Cary; Cleghorn, Tim; Cucinotta, Frank; Saganti, Premkumar; Atwell, William; Turner, Ron

2006-01-01

The helium energy spectrum in Martian orbit has been observed by the MARIE charged particle spectrometer aboard the 2001 Mars Odyssey spacecraft. The orbital data were taken from March 13, 2002 to October 28, 2003, at which time a very intense Solar Particle Event caused a loss of communication between the instrument and the spacecraft. The silicon detector stack in MARIE is optimized for the detection of protons and helium in the energy range below 100MeV/n, which typically includes almost all of the flux during SPEs. This also makes MARIE an efficient detector for GCR helium in the energy range of 50-150MeV/n. We will present the first fully normalized flux results from MARIE, using helium ions in this energy range

Helium cosmic ray flux measurements at Mars

Energy Technology Data Exchange (ETDEWEB)

Lee, Kerry [University of Houston, 4800 Calhoun Rd. Houston, TX 77204 (United States)]. E-mail: ktlee@ems.jsc.nasa.gov; Pinsky, Lawrence [University of Houston, 4800 Calhoun Rd. Houston, TX 77204 (United States); Andersen, Vic [University of Houston, 4800 Calhoun Rd. Houston, TX 77204 (United States); Zeitlin, Cary [National Space Biomedical Research Institute, Baylor College of Medicine, Houston, TX (United States); Cleghorn, Tim [NASA Johnson Space Center, 2101 NASA Road 1, Houston, TX 77058 (United States); Cucinotta, Frank [NASA Johnson Space Center, 2101 NASA Road 1, Houston, TX 77058 (United States); Saganti, Premkumar [Prairie View A and M University, P.O. Box 519, Prairie View, TX 77446-0519 (United States); Atwell, William [The Boeing Company, Houston, TX (United States); Turner, Ron [Advancing National Strategies and Enabling Results (ANSER), Arlington, Virginia (United States)

2006-10-15

The helium energy spectrum in Martian orbit has been observed by the MARIE charged particle spectrometer aboard the 2001 Mars Odyssey spacecraft. The orbital data were taken from March 13, 2002 to October 28, 2003, at which time a very intense Solar Particle Event caused a loss of communication between the instrument and the spacecraft. The silicon detector stack in MARIE is optimized for the detection of protons and helium in the energy range below 100MeV/n, which typically includes almost all of the flux during SPEs. This also makes MARIE an efficient detector for GCR helium in the energy range of 50-150MeV/n. We will present the first fully normalized flux results from MARIE, using helium ions in this energy range.

Atomistic simulation of helium bubble nucleation in palladium

Energy Technology Data Exchange (ETDEWEB)

Wang Liang [Department of Applied Physics, Hunan University, Changsha 410082 (China); Hu, Wangyu [Department of Applied Physics, Hunan University, Changsha 410082 (China)], E-mail: wangyuhu2001cn@yahoo.com.cn; Xiao Shifang [Department of Applied Physics, Hunan University, Changsha 410082 (China)], E-mail: sfxiao@yahoo.com.cn; Yang Jianyu [Department of Maths and Physics, Hunan Institute of Engineering, Xiangtan 411104 (China); Deng Huiqiu [Department of Applied Physics, Hunan University, Changsha 410082 (China)

2009-09-15

A palladium crystal has been constructed with 11808 atoms. 55 helium atoms occupied the octahedral position of palladium crystal are introduced and retained in a spherical region. Molecular dynamic simulations are performed in a constant temperature and constant volume ensemble (NVT) with temperature controlled by Nose-Hoover thermostat. The interactions between palladium atoms are described with modified analytic embedded atom method (MAEAM), the interactions between palladium atom and helium atom are in the form of Morse potential, and the interactions between helium atoms are in the form of L-J potential function. With the analysis of the radial distribution function (RDF) and microstructure, it reveals that some of helium atoms form a series of clusters with different size, and the nucleation core is random at low temperature, and which is the embryo of helium bubble. Increasing temperature can accelerate the process of bubble nucleation, and the clusters will aggregate and coalesce into a bigger one in which there are no palladium atoms, and it is considered as a helium bubble.

Helium-Charged La-Ni-Al Thin Films Deposited by Magnetron Sputtering

International Nuclear Information System (INIS)

Shi Liqun; Chen Deming; Xu Shilin; Liu Chaozhu; Hao Wanli; Zhou Zhuyin

2005-01-01

An advanced implantation of low energy helium-4 atoms during the La-Ni-Al film growth by adopting magnetron sputtering with Ar/He mixture gases is discussed. Both proton backscattering spectroscopy (PBS) and elastic recoil detection (ERD) analyses were adopted to measure helium concentration of the films and distribution in the near-surface region. Helium atoms with a high concentration incorporate evenly in deposited film. The introduction of the helium with no extra irradiation damage is expected by choosing suitable deposition conditions. It was found that amorphous and crystalline LaNi 5 -type structures can be achieved when sputtered with pure Ar and Ar/He mixture gases at room temperature, respectively. Thermal desorption experiments proposes that a part of hydrogen atoms are bound to trapped helium at crystal and releases together with helium. Only a small fraction of helium is released from the helium-vacancy clusters in lower temperature range and most of helium is released from small size helium bubbles in the high temperature range

Evolution of helium stars: a self-consistent determination of the boundary of a helium burning convective core

International Nuclear Information System (INIS)

Savonije, G.J.; Takens, R.J.

1976-01-01

A generalization of the Henyey-scheme is given that introduces the mass of the convective core and the density at the outer edge of the convective core boundary as unknowns which have to be solved simultaneously with the other unknowns. As a result, this boundary is determined in a physically self-consistent way for expanding as well as contracting cores, i.e. during the Henyey iterative cycle; its position becomes consistent with the overall physical structure of the star, including the run of the chemical abundances throughout the star. Using this scheme, the evolution of helium stars was followed up to carbon ignition for a number of stellar masses. As compared with some earlier investigations, the calculations show a rather large increase in mass of the convective cores during core helium burning. Evolutionary calculations for a 2M(sun) helium star show that the critical mass for which a helium star ignites carbon non-degenerately lies near 2M(sun). (orig.) [de

DIRECT EVALUATION OF THE HELIUM ABUNDANCES IN OMEGA CENTAURI

Energy Technology Data Exchange (ETDEWEB)

Dupree, A. K.; Avrett, E. H., E-mail: dupree@cfa.harvard.edu, E-mail: eavrett@cfa.harvard.edu [Harvard-Smithsonian Center for Astrophysics, Cambridge, MA 02138 (United States)

2013-08-20

A direct measure of the helium abundances from the near-infrared transition of He I at 1.08 {mu}m is obtained for two nearly identical red giant stars in the globular cluster Omega Centauri. One star exhibits the He I line; the line is weak or absent in the other star. Detailed non-local thermal equilibrium semi-empirical models including expansion in spherical geometry are developed to match the chromospheric H{alpha}, H{beta}, and Ca II K lines, in order to predict the helium profile and derive a helium abundance. The red giant spectra suggest a helium abundance of Y {<=} 0.22 (LEID 54064) and Y = 0.39-0.44 (LEID 54084) corresponding to a difference in the abundance {Delta}Y {>=} 0.17. Helium is enhanced in the giant star (LEID 54084) that also contains enhanced aluminum and magnesium. This direct evaluation of the helium abundances gives observational support to the theoretical conjecture that multiple populations harbor enhanced helium in addition to light elements that are products of high-temperature hydrogen burning. We demonstrate that the 1.08 {mu}m He I line can yield a helium abundance in cool stars when constraints on the semi-empirical chromospheric model are provided by other spectroscopic features.

Capacity enhancement of indigenous expansion engine based helium liquefier

Science.gov (United States)

Doohan, R. S.; Kush, P. K.; Maheshwari, G.

2017-02-01

Development of technology and understanding for large capacity helium refrigeration and liquefaction at helium temperature is indispensable for coming-up projects. A new version of helium liquefier designed and built to provide approximately 35 liters of liquid helium per hour. The refrigeration capacity of this reciprocating type expansion engine machine has been increased from its predecessor version with continuous improvement and deficiency debugging. The helium liquefier has been built using components by local industries including cryogenic Aluminum plate fin heat exchangers. Two compressors with nearly identical capacity have been deployed for the operation of system. Together they consume about 110 kW of electric power. The system employs liquid Nitrogen precooling to enhance liquid Helium yield. This paper describes details of the cryogenic expander design improvements, reconfiguration of heat exchangers, performance simulation and their experimental validation.

Liquid-helium scintillation detection with germanium photodiodes

International Nuclear Information System (INIS)

Luke, P.N.; Haller, E.E.; Steiner, H.M.

1982-05-01

Special high-purity germanium photodiodes have been developed for the direct detection of vacuum ultraviolet scintillations in liquid helium. The photodiodes are immersed in the liquid helium, and scintillations are detected through one of the bare sides of the photodiodes. Test results with scintillation photons produced by 5.3-MeV α particles are presented. The use of these photodiodes as liquid-helium scintillation detectors may offer substantial improvements over the alternate detection method requiring the use of wavelength shifters and photomultiplier tubes

Helium release from metals with face-centered cubic structure

International Nuclear Information System (INIS)

Sciani, V.; Lucki, G.; Jung, P.

1984-01-01

The helium release from gold sheets of 5 and 54 μm of thickness and helium concentrations from 10 -9 to 10 -5 ap of He during the isothermal and linear annealing is studied. The helium was put in the sample through the implantation of alpha particles, with variable energy,in the cyclotron. The free diffusion of the atoms of the helium, where the diffusion coefficient follows an Arrhenius law is studied. (E.G.) [pt

Present status of the disk pressure tests for hydrogen embrittlements

International Nuclear Information System (INIS)

Fidelle, J.P.

1988-01-01

The Disk Pressure Tests (DPT) have been developed considerably. Theoretically: a finite elements mechanical analysis shows the build up of a triaxial stress state already at the beginning of the test, which, with other reasons accounts for the very high sensitivity. Experimentally: for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for environment embrittlement due to H 2 hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably: up to 300 MPa and up to 1000 0 C. Very low strain rate - longer than a month - tests have been able to evidence HGE; of FCC alloys where H diffusivity is low for very oxidation -sensitive metals such as Nb and Ta, effects may appear only at somewhat high rates. The relationship between dynamic tests, static and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analysed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 15O + materials in different conditions. Comparison of HGE tests and service behaviour of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service, which, provided the shape of the stress strain curves is not significantly affected, can be expanded to IHE, HE by other fluids than H 2 , 36 refs

Feasibility of lunar Helium-3 mining

Science.gov (United States)

Kleinschneider, Andreas; Van Overstraeten, Dmitry; Van der Reijnst, Roy; Van Hoorn, Niels; Lamers, Marvin; Hubert, Laurent; Dijk, Bert; Blangé, Joey; Hogeveen, Joel; De Boer, Lennaert; Noomen, Ron

With fossil fuels running out and global energy demand increasing, the need for alternative energy sources is apparent. Nuclear fusion using Helium-3 may be a solution. Helium-3 is a rare isotope on Earth, but it is abundant on the Moon. Throughout the space community lunar Helium-3 is often cited as a major reason to return to the Moon. Despite the potential of lunar Helium-3 mining, little research has been conducted on a full end-to-end mission. This abstract presents the results of a feasibility study conducted by students from Delft University of Technology. The goal of the study was to assess whether a continuous end-to-end mission to mine Helium-3 on the Moon and return it to Earth is a viable option for the future energy market. The set requirements for the representative end-to-end mission were to provide 10% of the global energy demand in the year 2040. The mission elements have been selected with multiple trade-offs among both conservative and novel concepts. A mission architecture with multiple decoupled elements for each transportation segment (LEO, transfer, lunar surface) was found to be the best option. It was found that the most critical element is the lunar mining operation itself. To supply 10% of the global energy demand in 2040, 200 tons of Helium-3 would be required per year. The resulting regolith mining rate would be 630 tons per second, based on an optimistic concentration of 20 ppb Helium-3 in lunar regolith. Between 1,700 to 2,000 Helium-3 mining vehicles would be required, if using University of Wisconsin’s Mark III miner. The required heating power, if mining both day and night, would add up to 39 GW. The resulting power system mass for the lunar operations would be in the order of 60,000 to 200,000 tons. A fleet of three lunar ascent/descent vehicles and 22 continuous-thrust vehicles for orbit transfer would be required. The costs of the mission elements have been spread out over expected lifetimes. The resulting profits from Helium

Simplicity works for superfluid helium

International Nuclear Information System (INIS)

Bowley, Roger

2000-01-01

The famous philosopher Karl Popper once said that ''science is the art of systematic oversimplification''. Indeed, when faced with a new puzzle the trick is to simplify it without losing the essential physics - something that is easier said than done. However, this approach has paid off recently in low-temperature physics. Last year Richard Packard, Seamus Davis and co-workers at the University of California at Berkeley encountered a puzzling new phenomenon in superfluid helium-3, a quantum fluid that remains a liquid close to absolute zero and exhibits unusual properties such as the ability to flow without friction (A Machenkov et al. 1999 Phys. Rev. Lett. 83 3860). Previous experiments had revealed that certain effects in liquid helium are analogous to effects observed in superconductors, materials that lose all resistance to electric current at low temperatures. When the Berkeley researchers connected two reservoirs of superfluid helium-3, the superfluid flowed back and forth through apertures that formed a ''weak link'' between the two containers. This behaviour is similar to the oscillatory current of electrons that can flow across an insulating gap separating two superconductors - a device that is known as a Josephson junction. What was puzzling about the Berkeley results was that the helium-3 had two different stable configurations, both of which behaved in an unconventional way compared with a Josephson junction. This puzzle has now been solved independently by Sidney Yip at the National Center for Theoretical Sciences in Taiwan, and by Janne Viljas and Erkki Thuneberg at the Helsinki University of Technology in Finland (Phys. Rev. Lett. 1999 83 3864 and 3868). In this article the author describes the latest research on superfluid helium. (UK)

KSTAR Helium Refrigeration System Design and Manufacturing

International Nuclear Information System (INIS)

Dauguet, P.; Briend, P.; Abe, I.; Fauve, E.; Bernhardt, J.-M.; Andrieu, F.; Beauvisage, J.

2006-01-01

The tokamak developed in the KSTAR (Korean Superconducting Tokamak Advanced Research) project makes intensive use of superconducting magnets operated at 4.5 K. The cold components of the KSTAR tokamak require forced flow of supercritical helium for magnets/structure, boiling liquid helium for current leads, and gaseous helium for thermal shields. The cryogenic system will provide stable operation and full automatic control. A three-pressure helium cycle composed of six turbines has been customised design for this project. The '' design '' operating mode results with a system composed of a 9 kW refrigerator (including safety margin) and using gas and liquid storages for mass balancing. During Shot/Standby mode, the heat loads are highly time-dependent. A thermal damper is used to smooth these variations and will allow stable operation. (author)

Pierre Gorce working on a helium pump.

CERN Multimedia

1975-01-01

This type of pump was designed by Mario Morpurgo, to circulate liquid helium in superconducting magnets wound with hollow conductors. M. Morpurgo, Design and construction of a pump for liquid helium, CRYIOGENICS, February 1977, p. 91

Helium solubility and bubble growth in metals under high pressure

International Nuclear Information System (INIS)

Laakmann, J.

1985-07-01

Helium solubility and bubble growth in metals under high pressure polycrystals and single crystals of gold were heated in helium at temperatures between 475 K and 1250 K in a pressure regime of 200 to 2700 bar to measure the solubility of helium in gold. After quenching to room temperature the helium content, measured by mass spectrometry, showed the following properties: 1) A linear dependence of the He solubility on pressure. 2) Thinning of the specimen reduces the helium content by a factor 10 to 100 but does not change the linear pressure dependence. 3) The thermal release of He from thinned polycrystals and single crystals occurs mainly in a single peak at 500 K. 4) The He concentration of the thinned single crystals was lower by a factor of 10 to 50 than that of the thinned polycrystals. 5) The He solubility in single crystals can be described by an enthalpy of solution Hsub(s)sup(f) = 0.85 +- 0.7 eV and a non-configurational entropy of Ssub(s)sup(f) between 0 k and 1 k (k: Boltzmann-constant). In order to measure the pressure dependence of helium bubble growth in nickel polycrystal Ni-foils were α-implanted to a helium content of 130 appm. The evaluation of the size distribution of the helium bubbles after heat treatments shows 1) The helium content of the observable bubbles - assumed to be in equilibrium - equals the amount of helium implanted into the specimen. 2) The activation energy for the growth of helium bubbles is 1.25 +- 0.3 eV. The comparison of specimen which had been heated at low pressures up to 10 bar with others heated at 2500-2700 bar does not show an unequivocal pressure dependence for helium bubble growth. (orig./IHOE) [de

Irradiation embrittlement and mitigation. V. 1. Working material. Proceedings of a specialists meeting held in Espoo, Finland 23-26 October 1995

International Nuclear Information System (INIS)

1995-01-01

The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility experience on radiation damage and its surveillance, annealing and re-embrittlement of PWR, WWER and BWR reactor pressure vessel materials. The scope included: mechanism of radiation damage; effects of operating parameters (flux, temperature, time, etc.); results from surveillance programmes and their analysis; fracture mechanics testing and evaluation; annealing and optimization of the process; re-embrittlement after annealing. Presentations were aimed at better understanding of radiation damage, annealing and re-irradiation behaviour of reactor pressure vessels materials, at providing guidance and recommendations for optimization of annealing and surveillance programmes and directions for further investigations. Refs, figs and tabs

Irradiation embrittlement and mitigation. V. 1. Working material. Proceedings of a specialists meeting held in Espoo, Finland 23-26 October 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility experience on radiation damage and its surveillance, annealing and re-embrittlement of PWR, WWER and BWR reactor pressure vessel materials. The scope included: mechanism of radiation damage; effects of operating parameters (flux, temperature, time, etc.); results from surveillance programmes and their analysis; fracture mechanics testing and evaluation; annealing and optimization of the process; re-embrittlement after annealing; Presentations were aimed at better understanding of radiation damage, annealing and re-irradiation behaviour of reactor pressure vessels materials, at providing guidance and recommendations for optimization of annealing and surveillance programmes and directions for further investigations. Refs, figs and tabs.

Comment on theories for helium-assisted void nucleation

International Nuclear Information System (INIS)

Russell, K.C.

1976-01-01

Voids form by agglomeration of irradiation-induced vacancies which remain after preferential absorption of self interstitials at dislocation lines. Helium which is formed by (n,α) transmutations and, in simulation studies, may be ion-implanted, often plays an important, but puzzling role. In some materials, very few voids form in the absence of helium, even after intense irradiation. In many other materials , voids form readily under a variety of irradiation conditions, even in the absence of helium. Why some materials require helium - typically in the 10 -6 apa (atom per atom) range - and others do not, and the reason for that particular level are by no means clear. The physics of void nucleation, particularly the role of helium, have been the subject of several theoretical papers. This note presents a critique of these theories, and then briefly outlines a new analysis which is not subject to their limitations. (Auth.)

PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, B.J. [Oak Ridge National Lab., TN (United States)

1994-01-01

Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

Adsorption pump for helium pumping out

International Nuclear Information System (INIS)

Donde, A.L.; Semenenko, Yu.E.

1981-01-01

Adsorption pump with adsorbent cooling by liquid helium is described. Shuttered shield protecting adsorbent against radiation is cooled with evaporating helium passing along the coil positioned on the shield. The pump is also equipped with primed cylindrical shield, cooled with liquid nitrogen. The nitrogen shield has in the lower part the shuttered shield, on the pump casing there is a valve used for pump pre-burning, and valves for connection to recipient as well. Pumping- out rates are presented at different pressures and temperatures of adsorbent. The pumping-out rate according to air at absorbent cooling with liquid nitrogen constituted 5x10 -4 Pa-3000 l/s, at 2x10 -2 Pa-630 l/s. During the absorbent cooling with liquid hydrogen the pumping-out rate according to air was at 4x10 -4 Pa-580 l/s, at 2x10 -3 Pa-680 l/s, according to hydrogen - at 8x10 -5 Pa-2500 l/s, at 5x10 -3 Pa-4200 l/s. During adsorbent cooling with liquid helium the rate of pumping-out according to hydrogen at 3x10 5 Pa-2400% l/s, at 6x10 3 Pa-1200 l/s, and according to helium at 3.5x10 -5 Pa-2800 l/s, at 4x10 -3 Pa-1150 l/s. The limit vacuum is equal to 1x10 -7 Pa. The volume of the vessel with liquid helium is equal to 3.5 l. Helium consumption is 80 cm 3 /h. Consumption of liquid nitrogen from the shield is 400 cm 3 /h. The limit pressure in the pump is obtained after forevacuum pumping-out (adsorbent regeneration) at 300 K temperature. The pump is made of copper. The pump height together with primed tubes is 800 mm diameter-380 mm [ru

Blackbody-radiation correction to the polarizability of helium

International Nuclear Information System (INIS)

Puchalski, M.; Jentschura, U. D.; Mohr, P. J.

2011-01-01

The correction to the polarizability of helium due to blackbody radiation is calculated near room temperature. A precise theoretical determination of the blackbody radiation correction to the polarizability of helium is essential for dielectric gas thermometry and for the determination of the Boltzmann constant. We find that the correction, for not too high temperature, is roughly proportional to a modified hyperpolarizability (two-color hyperpolarizability), which is different from the ordinary hyperpolarizability of helium. Our explicit calculations provide a definite numerical result for the effect and indicate that the effect of blackbody radiation can be excluded as a limiting factor for dielectric gas thermometry using helium or argon.

Atmospheric helium and geomagnetic field reversals.

Science.gov (United States)

Sheldon, W. R.; Kern, J. W.

1972-01-01

The problem of the earth's helium budget is examined in the light of recent work on the interaction of the solar wind with nonmagnetic planets. It is proposed that the dominant mode of helium (He4) loss is ion pumping by the solar wind during geomagnetic field reversals, when the earth's magnetic field is very small. The interaction of the solar wind with the earth's upper atmosphere during such a period is found to involve the formation of a bow shock. The penetration altitude of the shock-heated solar plasma is calculated to be about 700 km, and ionization rates above this level are estimated for a cascade ionization (electron avalanche) process to average 10 to the 9th power ions/sq cm/sec. The calculated ionization rates and the capacity of the solar wind to remove ionized helium (He4) from the upper atmosphere during geomagnetic dipole reversals are sufficient to yield a secular equilibrium over geologic time scales. The upward transport of helium from the lower atmosphere under these conditions is found to be adequate to sustain the proposed loss rate.

Diffusion of helium in the Sun

Energy Technology Data Exchange (ETDEWEB)

Noerdlinger, P D [Michigan State Univ., East Lansing (USA). Dept. of Astronomy and Astrophysics; Amsterdam Univ. (Netherlands). Sterrenkundig Instituut)

1977-05-01

I have reduced the set of diffusion and flow equations developed by Burgers for a multi-component gas to a workable scheme for the actual evaluation of the relative diffusion of hydrogen and helium in stars. Previous analyses have used the Aller and Chapman equations, which apply only to trace constitutents and whose coefficients are not believed to be as accurate as Burgers'. Furthermore, the resulting equations have been combined consistently with Paczynski's stellar evolution code to demonstrate small but significant effects in the Sun, from the thermal and gravitational settling of Helium. The core helium content of a 1 M star goes up about 0.04 and the surface helium content down by about -0.03 in 4.5 10/sup 9/ years. The results are still somewhat uncertain because of uncertainties in the underlying plasma physics, and further research is suggested. In any case, the diffusion process speeds up with time, due to increased temperature gradient, and it will be of interest to follow the process in older stars and in later stellar evolution.

Antiprotonic helium atomcules

Directory of Open Access Journals (Sweden)

Sauge Sébastien

2012-10-01

Full Text Available About 3% of antiprotons ( stopped in helium are long-lived with microsecond lifetimes, against picoseconds in all other materials. This unusual longevity has been ascribed to the trapping of on metastable bound states in He+ helium atom-molecules thus named atomcules. Apart from their unique dual structure investigated by laser spectroscopy – a near-circular quasi-classical Rydberg atom with l ~ n – 1 ~ 37 or a special diatomic molecule with a negatively charged nucleus in high rotational state with J = l – the chemical physics aspects of their interaction with other atoms or molecules constitute an interesting topic for molecular physics. While atomcules may resist to million collisions in helium, molecular contaminants such as H2 are likely to destroy them in a single one, down to very low temperatures. In the Born-Oppenheimer framework, we interpret the molecular interaction obtained by ab initio quantum chemical calculations in terms of classical reactive channels, with activation barriers accounting for the experiments carried out in He and H2. From classical trajectory Monte Carlo simulations, we show that the thermalization stage strongly quenches initial populations, thus reduced to a recovered 3 % trapping fraction. This work illustrates the pertinence of chemical physics concepts to the study of exotic processes involving antimatter. New insights into the physico-chemistry of cold interstellar radicals are anticipated.

Superfluid helium-4: An introductory review

International Nuclear Information System (INIS)

Vinen, W.F.

1983-01-01

Helium was first liquefied by Kamerlingh Onnes in Leiden in July 1908, an achievement that followed much careful and painstaking work. On the same day Onnes reduced the temperature of his helium to a value approaching lK, and he must therefore have produced and observed the superfluid phase. These experimental discoveries led very quickly to a series of remarkable theoretical contributions that laid the foundations for all subsequent work. The period since the second world war has of course seen an enormous amount of work on superfluid helium-4. In reviewing it the author tries to see it in terms of two threads: one originating from Landau; the other from London

Realization of mechanical rotation in superfluid helium

Science.gov (United States)

Gordon, E. B.; Kulish, M. I.; Karabulin, A. V.; Matyushenko, V. I.; Dyatlova, E. V.; Gordienko, A. S.; Stepanov, M. E.

2017-09-01

The possibility of using miniaturized low-power electric motors submerged in superfluid helium for organization of rotation inside a cryostat has been investigated. It has been revealed that many of commercial micromotors can operate in liquid helium consuming low power. Turret with 5 sample holders, assembled on the base of stepper motor, has been successfully tested in experiments on the nanowire production in quantized vortices of superfluid helium. Application of the stepper motor made it possible in a single experiment to study the effect of various experimental parameters on the yield and quality of the nanowires. The promises for continuous fast rotation of the bath filled by superfluid helium by using high-speed brushless micromotor were outlined and tested. Being realized, this approach will open new possibility to study the guest particles interaction with the array of parallel linear vortices in He II.

Helium production in mixed spectrum reactor-irradiated pure elements

International Nuclear Information System (INIS)

Kneff, D.W.; Oliver, B.M.; Skowronski, R.P.

1986-01-01

The objectives of this work are to apply helium accumulation neutron dosimetry to the measurement of neutron fluences and energy spectra in mixed-spectrum fission reactors utilized for fusion materials testing, and to measure helium generation rates of materials in these irradiation environments. Helium generation measurements have been made for several Fe, Cu Ti, Nb, Cr, and Pt samples irradiated in the mixed-spectrum High Flux Isotope Reactor (HFIR) and Oak Ridge Research Reactor (ORR) at the Oak Ridge National Laboratory. The results have been used to integrally test the ENDF/B-V Gas Production File, by comparing the measurements with helium generation predictions made by Argonne National Laboratory using ENDF/B-V cross sections and adjusted reactor spectra. The comparisons indicate consistency between the helium measurements and ENDF/B-V for iron, but cross section discrepancies exist for helium production by fast neutrons in Cu, Ti, Nb, and Cr (the latter for ORR). The Fe, Cu, and Ti work updates and extends previous measurements

Interdiffusion of krypton and xenon in high-pressure helium

International Nuclear Information System (INIS)

Campana, R.J.; Jensen, D.D.; Epstein, B.D.; Hudson, R.G.; Baldwin, N.L.

1980-01-01

The interdiffusion of gaseous fission products in high-pressure helium is an important factor in the control of radioactivity in gas-cooled fast breeder reactors (GCFRs). As presently conceived, GCFRs use pressure-equalized and vented fuel in which fission gases released from the solid matrix oxide fuel are transported through the fuel rod interstices and internal fission product traps to the fuel assembly vents, where they are swept away to external traps and storage. Since the predominant transport process under steady-state operating conditions is interdiffusion of gaseous fission products in helium, the diffusion properties of krypton-helium and xenon-helium couples have been measured over the range of GCFR temperature and pressure conditions ( -1 ) and expected temperature dependence to the 1.66 power (Tsup(1.66)) at lower pressures and temperatures. Additional work is in progress to measure the behaviour of the krypton-helium and xenon-helium couples in GCFR fuel rod charcoal delay traps. (author)

Simplicity works for superfluid helium

Energy Technology Data Exchange (ETDEWEB)

Bowley, Roger [University of Nottingham, Nottingham (United Kingdom)

2000-02-01

The famous philosopher Karl Popper once said that ''science is the art of systematic oversimplification''. Indeed, when faced with a new puzzle the trick is to simplify it without losing the essential physics - something that is easier said than done. However, this approach has paid off recently in low-temperature physics. Last year Richard Packard, Seamus Davis and co-workers at the University of California at Berkeley encountered a puzzling new phenomenon in superfluid helium-3, a quantum fluid that remains a liquid close to absolute zero and exhibits unusual properties such as the ability to flow without friction (A Machenkov et al. 1999 Phys. Rev. Lett. 83 3860). Previous experiments had revealed that certain effects in liquid helium are analogous to effects observed in superconductors, materials that lose all resistance to electric current at low temperatures. When the Berkeley researchers connected two reservoirs of superfluid helium-3, the superfluid flowed back and forth through apertures that formed a ''weak link'' between the two containers. This behaviour is similar to the oscillatory current of electrons that can flow across an insulating gap separating two superconductors - a device that is known as a Josephson junction. What was puzzling about the Berkeley results was that the helium-3 had two different stable configurations, both of which behaved in an unconventional way compared with a Josephson junction. This puzzle has now been solved independently by Sidney Yip at the National Center for Theoretical Sciences in Taiwan, and by Janne Viljas and Erkki Thuneberg at the Helsinki University of Technology in Finland (Phys. Rev. Lett. 1999 83 3864 and 3868). In this article the author describes the latest research on superfluid helium. (UK)

Application of vanadium alloys to a fusion reactor blanket

Energy Technology Data Exchange (ETDEWEB)

Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

1984-05-01

Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

Surveillance as a complement to irradiation embrittlement studies: Status and needs

International Nuclear Information System (INIS)

Steele, L.E.

1977-01-01

The history of the study of radiation embrittlement of reactor pressure vessel steels has gone through three stages in the USA. 1) A scientific curiosity. 2) Empirical or laboratory evaluation of typical steels, and 3) Integration of the scientific and empirical to advance status and evolve standard techniques. The current stage is one in which surveillance data compliments the laboratory studies which characterized Stage 3. The early USA surveillance programs were generally analyzed by the same people who were the primary laboratory investigators. An effort must be made to continue this type of collaboration as a useful two-way learning procedure though it will become more and more difficult as nuclear power is broadly commercialized. The current status of both types of USA programs will be presented to encourage the most advantageous use of data from both sources. At this time about 25 USA nuclear power reactors have operated long enough to have provided initial surveillance or dosimetry results. An effort will be made to summarize the general status of these in order to: 1) Provide complimentary data to laboratory studies. 2) Assess directions in handling the problems of radiation embrittlement. 3) Note lessons learned for improving surveillance efforts in the future. 4) Identify possible research tasks for the future to support in-service surveillance and other measures. 5) Justify facts advancing surveillance requirements to status of national codes and standards. 6) Justify facts requiring changes in current national codes and standards. A plan will be presented along with an introduction of each member of the USA delegation for systematic presentation of the status of reactor vessel surveillance in the USA. (author)

Helium turbomachine design for GT-MHR power plant

International Nuclear Information System (INIS)

McDonald, C.F.; Orlando, R.J.

1994-07-01

The power conversion system in the gas turbine modular helium reactor (GT-MHR) power plant is based on a highly recuperated closed Brayton cycle. The major component in the direct cycle system is a helium closed-cycle gas turbine rated at 286 MW(e). The rotating group consists of an intercooled helium turbocompressor coupled to a synchronous generator. The vertical rotating assembly is installed in a steel vessel, together with the other major components (i.e., recuperator, precooler, intercooler, and connecting ducts and support structures). The rotor is supported on an active magnetic bearing system. The turbine operates directly on the reactor helium coolant, and with a temperature of 850 degree C (1562 degree F) the plant efficiency is over 47%. This paper addresses the design and development planning of the helium turbomachine, and emphasizes that with the utilization of proven technology, this second generation nuclear power plant could be in service in the first decade of the 21st century

Optomechanics in a Levitated Droplet of Superfluid Helium

Science.gov (United States)

Brown, Charles; Harris, Glen; Harris, Jack

2017-04-01

A critical issue common to all optomechanical systems is dissipative coupling to the environment, which limits the system's quantum coherence. Superfluid helium's extremely low optical and mechanical dissipation, as well as its high thermal conductivity and its ability cool itself via evaporation, makes the mostly uncharted territory of superfluid optomechanics an exciting avenue for exploring quantum effects in macroscopic objects. I will describe ongoing work that aims to exploit the unique properties of superfluid helium by constructing an optomechanical system consisting of a magnetically levitated droplet of superfluid helium., The optical whispering gallery modes (WGMs) of the droplet, as well as the mechanical oscillations of its surface, should offer exceptionally low dissipation, and should couple to each other via the usual optomechanical interactions. I will present recent progress towards this goal, and also discuss the background for this work, which includes prior demonstrations of magnetic levitation of superfluid helium, high finesse WGMs in liquid drops, and the self-cooling of helium drops in vacuum.

Status on the selection and development of an embrittlement trend curve to use in ASTM standard guide E900

International Nuclear Information System (INIS)

Kirk, M.; Brian Hall, J.; Server, W.; Lucon, E.; Erickson, M.; Stoller, R.

2015-01-01

ASTM E900-07, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, includes an embrittlement trend curve. The trend curve can be used to predict the effect of neutron irradiation on the embrittlement of ferritic pressure vessel steels, as quantified by the shift in the Charpy V-Notch transition curve at 41 Joules of absorbed energy (ΔT 41J ). The current E900 trend curve was first adopted in the 2002 revision. In 2011 ASTM Subcommittee E10.02 undertook an extensive effort to evaluate the adequacy of the E900 trend curve for continued use. This paper summarizes the current status of this effort, which has produced a trend curve calibrated using a database of over 1800 ΔT 41J values from the light water reactor surveillance programs in thirteen countries. (authors)

Metallic materials for the hydrogen energy industry and main gas pipelines: complex physical problems of aging, embrittlement, and failure

International Nuclear Information System (INIS)

Nechaev, Yu S

2008-01-01

The possibilities of effective solutions of relevant technological problems are considered based on the analysis of fundamental physical aspects, elucidation of the micromechanisms and interrelations of aging and hydrogen embrittlement of materials in the hydrogen industry and gas-main industries. The adverse effects these mechanisms and processes have on the service properties and technological lifetime of materials are analyzed. The concomitant fundamental process of formation of carbohydride-like and other nanosegregation structures at dislocations (with the segregation capacity 1 to 1.5 orders of magnitude greater than in the widely used Cottrell 'atmosphere' model) and grain boundaries is discussed, as is the way in which these structures affect technological processes (aging, hydrogen embrittlement, stress corrosion damage, and failure) and the physicomechanical properties of the metallic materials (including the technological lifetimes of pipeline steels). (reviews of topical problems)

Born-Mayer type molybdenum-helium and helium-helium interaction potentials, fitted to the results of the helium desorption experiments

International Nuclear Information System (INIS)

Heugten, W.F.W.M. van; Veen, A. van; Caspers, L.M.

1979-01-01

Classes of Born-Mayer type Mo-He and He-He potentials have been derived from helium desorption experiments. The classes are described by linear relations between the Born-Mayer parameters A and b. For computer simulations the Mo-He potential phisub(MoHe)(r)=exp (6.5-3.63 r) and the He-He potential phisub(HeHe)(r)=exp(5.3-5.51 r) are proposed. (Auth.)

The role of pressure vessel embrittlement in the long term operation of nuclear power plants

International Nuclear Information System (INIS)

Ballesteros, A.; Ahlstrand, R.; Bruynooghe, C.; Estorff, U. von; Debarberis, L.

2012-01-01

Highlights: ► Relevant open scientific issues for the long term operation of RPVs are discussed (flux effect, late blooming phases, etc.). ► Several European and American research programmes dealing with these open issues are reviewed. ► A method for consolidation and preservation of knowledge in this field is presented. - Abstract: The lack of new build of plants over the last twenty years has resulted in a switch within the industry from design, construction and development of new systems to the strengthening of safety systems and to the life extension, or long term operation (LTO), of existing reactors. The most relevant component of any nuclear power plan (NPP) is the reactor pressure vessel (RPV). This is because currently the RPV is still considered irreplaceable or prohibitively expensive to replace. This means, that if it degrades sufficiently, it could be the operational life limiting feature of the NPP. A RPV operational life of 60 years is being considered frequently by many utilities in their plant life management programmes. Areas of improvement facing long term operation are the reduction of uncertainties in the embrittlement parameters of irradiated vessels, and the development of embrittlement trend curves at high fluence levels, where surveillance data are scarce. Different techniques can be used to upgrade the surveillance programmes, as the use of miniature or reconstituted specimens and the application of best estimate assessment tools (e.g. Master Curve). Several relevant international research projects are on-going or have been proposed to clarify the material condition of long operated vessels. Knowledge management is a complementary tool, but not for it less important. The general context for LTO of RPVs is presented in this paper. Starting with a review of relevant embrittlement issues still open, followed by presenting the different techniques and tools that can be used to support LTO, and summarising the scopes of relevant European

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

International Nuclear Information System (INIS)

Krasikov, E. A.

2012-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

International Nuclear Information System (INIS)

Krasikov, E.A.

2012-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 o C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 o C and following extra irradiation (87 h at 330 o C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated management methods, will help

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

Energy Technology Data Exchange (ETDEWEB)

Krasikov, E. A. [National Research Centre Kurchatov Inst., 1, Kurchatov Sq., Moscow, 123182 (Russian Federation)

2012-07-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which

FP7 project LONGLIFE: Treatment of long-term irradiation embrittlement effects in RPV safety assessment

International Nuclear Information System (INIS)

May, J.; Hein, H.; Altstadt, E.; Bergner, F.; Viehrig, H.W.; Ulbricht, A.; Chaouadi, R.; Radiguet, B.; Cammelli, S.; Huang, H.; Wilford, K.

2012-01-01

The increasing age of European Nuclear Power Plants (NPPs) and envisaged extensions of plant lifetimes from 40 up to 80 years require an improved understanding of ageing phenomena of RPV components. The Network of Excellence NULIFE (Nuclear Plant Life Prediction) has been established to advance the safe and economic long-term operation (LTO) of NPPs by facilitating increased co-operation for applied R and D amongst members of the European nuclear community. The accurate prediction and management of RPV neutron irradiation embrittlement connected with long-term operation is an important aspect of this co-operation. Phenomena that might become important at high neutron fluences (such as flux effects and late blooming effects) have to be considered adequately in safety assessments. However, the surveillance database for prolonged irradiation times and low neutron fluxes is sparse. Consequently, there are significant uncertainties in the treatment of long-term irradiation effects. Therefore, the project LONGLIFE (Treatment of long-term irradiation embrittlement effects in RPV safety assessment) was initiated under the Euratom 7th Framework Programme of the European Commission as an umbrella project of NULIFE. LONGLIFE aims at 1) improved understanding of long-term irradiation phenomena that might compromise RPV integrity, and thereby the LTO of European NPPs, and 2) assessment of the adequacy of current prediction tools, codes, standards and surveillance guidelines for supporting long-term RPV operation. The scope of the work comprises the analysis of LTO boundary conditions; microstructural investigations and supplementary mechanical tests on RPV steels, including RPV steels from decommissioned plants; training activities; and elaboration of recommendations for RPV materials assessment and embrittlement surveillance under LTO conditions. A key part of the technical work is the selection of relevant materials for examination, e.g. which contain different weld and base

Neutral helium beam probe

Science.gov (United States)

Karim, Rezwanul

1999-10-01

This article discusses the development of a code where diagnostic neutral helium beam can be used as a probe. The code solves numerically the evolution of the population densities of helium atoms at their several different energy levels as the beam propagates through the plasma. The collisional radiative model has been utilized in this numerical calculation. The spatial dependence of the metastable states of neutral helium atom, as obtained in this numerical analysis, offers a possible diagnostic tool for tokamak plasma. The spatial evolution for several hypothetical plasma conditions was tested. Simulation routines were also run with the plasma parameters (density and temperature profiles) similar to a shot in the Princeton beta experiment modified (PBX-M) tokamak and a shot in Tokamak Fusion Test Reactor tokamak. A comparison between the simulation result and the experimentally obtained data (for each of these two shots) is presented. A good correlation in such comparisons for a number of such shots can establish the accurateness and usefulness of this probe. The result can possibly be extended for other plasma machines and for various plasma conditions in those machines.

Design and study of Engineering Test Facility - Helium Circulator

International Nuclear Information System (INIS)

Jiang Huijing; Ye Ping; Zhao Gang; Geng Yinan; Wang Jie

2015-01-01

Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability. (author)

Experimental Validation of the LHC Helium Relief System Flow Modeling

CERN Document Server

Fydrych, J; Riddone, G

2006-01-01

In case of simultaneous resistive transitions in a whole sector of magnets in the Large Hadron Collider, the helium would be vented from the cold masses to a dedicated recovery system. During the discharge the cold helium will eventually enter a pipe at room temperature. During the first period of the flow the helium will be heated intensely due to the pipe heat capacity. To study the changes of the helium thermodynamic and flow parameters we have simulated numerically the most critical flow cases. To verify and validate numerical results, a dedicated laboratory test rig representing the helium relief system has been designed and commissioned. Both numerical and experimental results allow us to determine the distributions of the helium parameters along the pipes as well as mechanical strains and stresses.

Severe Embrittlement of Neutron Irradiated Austenitic Steels Arising from High Void Swelling

International Nuclear Information System (INIS)

Neustroev, V.S.; Garner, F.

2007-01-01

Full text of publication follows: Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components. Similar loss of ductility is expected when swelling arises in fusion and light water reactor environments. Above 7-16% swelling there is complete loss of ductility, with the onset of ductility loss beginning at lower swelling in ring-pull tensile tests than for flat tensile specimens. For steels that develop extensive precipitation during irradiation, the critical swelling level is even lower. A model is presented to demonstrate the effect of voids acting alone to produce the embrittlement. Although voids are not very effective hardeners, they are very effective to generate stress concentrations between voids. The stress concentration ratio increases strongly when the void diameter exceeds ∼40% of the void-to-void separation distance. When the volume fraction of voids is rather high (about 16 % and higher), a geometric situation develops where it is possible to create an intense field of deformation glide planes residing at an angle of 45 deg. to the void-to-void axis. Significant localized flow then proceeds on these planes for specimen stress levels that are significantly lower than the yield stress. Voids also segregate nickel to their surfaces such that flow localization occurs in the low-nickel inter-void regions to produce strain-induced martensite, which is further accelerated by stress concentrations at the advancing crack tip, leading to catastrophic failure. (authors)

NUCLEAR CONDENSATE AND HELIUM WHITE DWARFS

Energy Technology Data Exchange (ETDEWEB)

Bedaque, Paulo F.; Berkowitz, Evan [Maryland Center for Fundamental Physics, Department of Physics, University of Maryland, College Park, MD (United States); Cherman, Aleksey, E-mail: bedaque@umd.edu, E-mail: evanb@umd.edu, E-mail: a.cherman@damtp.cam.ac.uk [Department of Applied Mathematics and Theoretical Physics, University of Cambridge, Cambridge CB3 0WA (United Kingdom)

2012-04-10

We consider a high-density region of the helium phase diagram, where the nuclei form a Bose-Einstein condensate rather than a classical plasma or a crystal. Helium in this phase may be present in helium-core white dwarfs. We show that in this regime there is a new gapless quasiparticle not previously noticed, arising when the constraints imposed by gauge symmetry are taken into account. The contribution of this quasiparticle to the specific heat of a white dwarf core turns out to be comparable in a range of temperatures to the contribution from the particle-hole excitations of the degenerate electrons. The specific heat in the condensed phase is two orders of magnitude smaller than in the uncondensed plasma phase, which is the ground state at higher temperatures, and four orders of magnitude smaller than the specific heat that an ion lattice would provide, if formed. Since the specific heat of the core is an important input for setting the rate of cooling of a white dwarf star, it may turn out that such a change in the thermal properties of the cores of helium white dwarfs has observable implications.

NUCLEAR CONDENSATE AND HELIUM WHITE DWARFS

International Nuclear Information System (INIS)

Bedaque, Paulo F.; Berkowitz, Evan; Cherman, Aleksey

2012-01-01

We consider a high-density region of the helium phase diagram, where the nuclei form a Bose-Einstein condensate rather than a classical plasma or a crystal. Helium in this phase may be present in helium-core white dwarfs. We show that in this regime there is a new gapless quasiparticle not previously noticed, arising when the constraints imposed by gauge symmetry are taken into account. The contribution of this quasiparticle to the specific heat of a white dwarf core turns out to be comparable in a range of temperatures to the contribution from the particle-hole excitations of the degenerate electrons. The specific heat in the condensed phase is two orders of magnitude smaller than in the uncondensed plasma phase, which is the ground state at higher temperatures, and four orders of magnitude smaller than the specific heat that an ion lattice would provide, if formed. Since the specific heat of the core is an important input for setting the rate of cooling of a white dwarf star, it may turn out that such a change in the thermal properties of the cores of helium white dwarfs has observable implications.

Development of an Agent-based Model to Analyze Contemporary Helium Markets

Energy Technology Data Exchange (ETDEWEB)

Riddle, Matthew E. [Argonne National Lab. (ANL), Argonne, IL (United States); Uckun, Canan [Argonne National Lab. (ANL), Argonne, IL (United States); Conzelmann, Guenter [Argonne National Lab. (ANL), Argonne, IL (United States); Macal, Charles M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-01

Although U.S. helium demand has remained relatively flat since 2009, exports of helium have increased significantly since then, driven primarily by demand for electronic and semiconductor manufacturing in Asia. In the midst of this global demand shift, the Helium Act dictates a new procedure for pricing and distributing the gas through a reserve that historically functioned as a loose “oligarchy.” The new procedure requires prices to be determined by the open market through auctions and a survey of market prices, as opposed to increasing prices according to the consumer price index. Response to these changes has caused temporary shortages, price increases, and a significant increase in the development of the helium extraction technologies used to produce helium from formerly marginal sources. Technologies are being developed and refined to extract helium from formerly low-yielding natural gas fields containing much lower amounts of helium than the previously considered economic threshold of 0.3%. Combining these transformative policies with the potential for new and significant global supplies from Qatar, Algeria, and Russia could lead to new and unforeseen market behaviors and reactions from global helium markets. The objective of the project is to analyze the global helium markets.

Organ protection by the noble gas helium

NARCIS (Netherlands)

Smit, K.F.

2017-01-01

The aims of this thesis were to investigate whether helium induces preconditioning in humans, and to elucidate the mechanisms behind this possible protection. First, we collected data regarding organ protective effects of noble gases in general, and of helium in particular (chapters 1-3). In chapter

Construction and testing of a double acting bellows liquid helium pump

International Nuclear Information System (INIS)

Burns, W.A.; Green, M.A.; Ross, R.R.; Van Slyke, H.

1980-05-01

The double acting reciprocating bellows liquid helium pump built and tested at the Lawrence Berkeley Laboratory is described. The pump is capable of delivering 50 gs -1 of liquid helium to supply the two-phase cooling sytem for a large superconducting magnet. The pump is driven by a torque motor at room temperature; the reciprocating motion is transmitted to the pump through a shaft which operates between room temperature and 4 0 K. The design details of this liquid helium pump are presented. The helium pump has operated in a helium bath and in pumped forced flow helium circuits. The results of these experimental tests are presented in this report

Helium leak testing the Westinghouse LCP coil

International Nuclear Information System (INIS)

Merritt, P.A.; Attaar, M.H.; Hordubay, T.D.

1983-01-01

The tests, equipment, and techniques used to check the Westinghouse LCP coil for coolant flow path integrity and helium leakage are unique in terms of test sensitivity and application. This paper will discuss the various types of helium leak testing done on the LCP coil as it enters different stages of manufacture. The emphasis will be on the degree of test sensitivity achieved under shop conditions, and what equipment, techniques and tooling are required to achieve this sensitivity (5.9 x 10 -8 scc/sec). Other topics that will be discussed are helium flow and pressure drop testing which is used to detect any restrictions in the flow paths, and the LCP final acceptance test which is the final leak test performed on the coil prior to its being sent for testing. The overall allowable leak rate for this coil is 5 x 10 -6 scc/sec. A general evaluation of helium leak testing experience are included

Development of helium isotopic database in Japan

International Nuclear Information System (INIS)

Kusano, Tomohiro; Asamori, Koichi; Umeda, Koji

2012-09-01

We constructed “Helium Isotopic Database in Japan”, which includes isotope ratios of noble gases and chemical compositions of gas samples collected from hot springs and drinking water wells. The helium isotopes are excellent natural tracers for indicating the presence of mantle derived volatiles, because they are chemically inert and thus conserved in crustal rock-water systems. It is common knowledge that mantle degassing does not occur homogeneously over the Earth's surface. The 3 He/ 4 He ratios higher than the typical crustal values are interpreted to indicate that transfer of mantle volatiles into the crust by processes or mechanisms such as magmatic intrusion, faulting. In particular the spatial variation of helium isotope ratios could provide a valuable information to identify volcanic regions and tectonically active areas. The database was compiled geochemical data of hot spring gas etc. from 108 published papers. As a result of the data compiling, the database has 1728 helium isotopic data. A CD-ROM is attached as an appendix. (author)

Hybrid Circuit QED with Electrons on Helium

Science.gov (United States)

Yang, Ge

Electrons on helium (eHe) is a 2-dimensional system that forms naturally at the interface between superfluid helium and vacuum. It has the highest measured electron mobility, and long predicted spin coherence time. In this talk, we will first review various quantum computer architecture proposals that take advantage of these exceptional properties. In particular, we describe how electrons on helium can be combined with superconducting microwave circuits to take advantage of the recent progress in the field of circuit quantum electrodynamics (cQED). We will then demonstrate how to reliably trap electrons on these devices hours at a time, at millikelvin temperatures inside a dilution refrigerator. The coupling between the electrons and the microwave resonator exceeds 1 MHz, and can be reproduced from the design geometry using our numerical simulation. Finally, we will present our progress on isolating individual electrons in such circuits, to build single-electron quantum dots with electrons on helium.

Formation of the lunar helium corona and atmosphere

Science.gov (United States)

Hodges, R. R., Jr.

1977-01-01

Helium is one of the dominant gases of the lunar atmosphere. Its presence is easily identified in data from the mass spectrometer at the Apollo 17 landing site. The major part of these data was obtained in lunar nighttime, where helium concentration reaches the maximum of its diurnal cyclic variation. The large night to day concentration ratio agrees with the basic theory of exospheric lateral transport reported by Hodges and Johnson (1968). A reasonable fraction of atmospheric helium atoms has a velocity in excess of the gravitational escape velocity. The result is a short average lifetime and a tenuous helium atmosphere. A description is presented of an investigation which shows that the atmosphere of the moon has two distinct components including low energy atoms, which are gravitationally bound in trajectories that intersect the lunar surface, and higher energy atoms, which are trapped in satellite orbits. The total helium abundance in the lunar corona is shown to be about 1.3 times 10 to the 30th power atoms.

Mechanical properties and microstructure of austenitic steels loaded with helium using tritium

International Nuclear Information System (INIS)

Sacovy, Paulette; Brun, Gilbert; Delaplace, Jean; Devaux, Joel; Fidelle, J.P.

1982-06-01

Following a review of the principle of the method using the radioactive decay of tritium to helium 3 to introduce helium into thick metallic test specimens, the results of preliminary tests performed on austenitic steels are presented. 304L and 316 steel specimens were loaded with helium, treated at 760 0 C to precipitate the helium in bubbles, and then strained by tensile stress at 800 0 C. In the material most loaded with helium (304 steel containing 230 ppm at. helium), a significant increase in distributed and total elongation at 800 0 C was observed. In the least loaded material (steel 316 containing 13 ppm at. helium), only total elongation decreased. Neither the yield stress nor the breaking load was altered by the presence of helium. Observations made by electron microscopy in the most loaded material revealed the presence, after heat treatment at 760 0 C, of very small bubbles of helium and also imperfect dislocation loops. The cause of these fault loops is discussed [fr

Orientation, alignment and polarisation in electron-helium collisions

International Nuclear Information System (INIS)

Beijers, J.P.M.

1987-01-01

In this thesis electron-photon coincidence experiments to study the excitation of helium by electron impact are updated. This is achieved by cross firing a well collimated and mono-energetic electron beam with a thermal helium beam and measuring the angular and/or polarisation distribution of the decay photons in coincidence with the inelastically scattered electrons. In this way target parameters are determined for the 2 1 P, 3 1 P, 3 1 D and 3 3 P states of helium. (Auth.)

Preservation and release dose of helium implanted in nanocrystal titanium film

International Nuclear Information System (INIS)

Long Xinggui; Luo Shunzhong; Peng Shuming; Zheng Sixiao; Liu Zhongyang; Wang Peilu; Liao Xiaodong; Liu Ning

2003-01-01

Helium concentration profile, preservation dose and release rate from a nanocrystal titanium film implanted with helium at an energy of 100 keV and dose of 2.2 x 10 18 cm -2 are measured by proton Rutherford backscattering technique in a range from room temperature to 400 degree C. The implanted helium may be stably preserved up to the 68 percent after keeping a long time of 210 d in the nanocrystal titanium film at the room temperature environment, and the He-Ti atomic ratio reaches to 52.6%. When the temperature of specimen increases to 100 degree C, the helium concentration can be preserved to 89.6% of the keeping helium dose at room temperature and He-Ti atomic ratio reaches 44%. Even if the specimen temperature up to 400 degree C, the helium concentration still can be preserved to 32.6% of the keeping helium dose at room temperature and the He-Ti atomic ratio is 17.1%. Possible mechanism of helium effectively preserved in the nanocrystal titanium film is discussed based on the energy stability viewpoint

Multiscale Modeling of Hydrogen Embrittlement for Multiphase Material

KAUST Repository

Al-Jabr, Khalid A.

2014-05-01

Hydrogen Embrittlement (HE) is a very common failure mechanism induced crack propagation in materials that are utilized in oil and gas industry structural components and equipment. Considering the prediction of HE behavior, which is suggested in this study, is one technique of monitoring HE of equipment in service. Therefore, multi-scale constitutive models that account for the failure in polycrystalline Body Centered Cubic (BCC) materials due to hydrogen embrittlement are developed. The polycrystalline material is modeled as two-phase materials consisting of a grain interior (GI) phase and a grain boundary (GB) phase. In the first part of this work, the hydrogen concentration in the GI (Cgi) and the GB (Cgb) as well as the hydrogen distribution in each phase, were calculated and modeled by using kinetic regime-A and C, respectively. In the second part of this work, this dissertation captures the adverse effects of hydrogen concentration, in each phase, in micro/meso and macro-scale models on the mechanical behavior of steel; e.g. tensile strength and critical porosity. The models predict the damage mechanisms and the reduction in the ultimate strength profile of a notched, round bar under tension for different hydrogen concentrations as observed in the experimental data available in the literature for steels. Moreover, the study outcomes are supported by the experimental data of the Fractography and HE indices investigation. In addition to the aforementioned continuum model, this work employs the Molecular Dynamics (MD) simulations to provide information regarding bond formulation and breaking. The MD analyses are conducted for both single grain and polycrystalline BCC iron with different amounts of hydrogen and different size of nano-voids. The simulations show that the hydrogen atoms could form the transmission in materials configuration from BCC to FCC (Face Centered Cubic) and HCP (Hexagonal Close Packed). They also suggest the preferred sites of hydrogen for

On the shear strength of tungsten nano-structures with embedded helium

International Nuclear Information System (INIS)

Smirnov, R.D.; Krasheninnikov, S.I.

2013-01-01

Modification of plastic properties of tungsten nano-structures under shear stress load due to embedded helium atoms is studied using molecular dynamics modelling. The modelling demonstrates that the yield strength of tungsten nano-structures reduces significantly with increasing embedded helium concentration. At high helium concentrations (>10 at%), the yield strength decreases to values characteristic to the pressure in helium nano-bubbles, which are formed in tungsten under such conditions and thought to be responsible for the formation of nano-fuzz on tungsten surfaces irradiated with helium plasma. It is also shown that tungsten plastic flow strongly facilitates coagulation of helium clusters to larger bubbles. The temperature dependencies of the yield strength are obtained. (letter)

Using Uncertainty Principle to Find the Ground-State Energy of the Helium and a Helium-like Hookean Atom

Science.gov (United States)

Harbola, Varun

2011-01-01

In this paper, we accurately estimate the ground-state energy and the atomic radius of the helium atom and a helium-like Hookean atom by employing the uncertainty principle in conjunction with the variational approach. We show that with the use of the uncertainty principle, electrons are found to be spread over a radial region, giving an electron…

Electrostatic charging and levitation of helium II drops

International Nuclear Information System (INIS)

Niemela, J.J.

1997-01-01

Liquid Helium II drops, of diameter 1 mm or less, are charged with positive helium ions and subsequently levitated by static electric fields. Stable levitation was achieved for drops of order 100-150 micrometers in diameter. The suspended drops could be translated to arbitrary positions within the levitator using additional superimposed DC electric fields, and also could be made to oscillate stably about their average positions by means of an applied time-varying electric field. A weak corona discharge was used to produce the necessary ions for levitation. A novel superfluid film flow device, developed for the controlled deployment of large charged drops, is described. Also discussed is an adjustable electric fountain that requires only a field emission tip operating at modest potentials, and works in both Helium I and Helium II

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-03-15

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

Five second helium neutral beam injection using argon-frost cryopumping techniques

International Nuclear Information System (INIS)

Phillips, J.C.; Kellman, D.H.; Hong, R.; Kim, J.; Laughon, G.M.

1995-01-01

High power helium neutral beams for the heating of tokamak discharges can now be provided for 5 s by using argon cryopumping (of the helium gas) in the beamlines. The DIII-D neutral beam system has routinely provided up to 20 MW of deuterium neutral beam heating in support of experiments on the DIII-D tokamak. Operation of neutral beams with helium has historically presented a problem in that pulse lengths have been limited to 500 ms due to reliance solely on volume pumping of the helium gas. Helium is not condensed on the cryopanels. A system has now been installed to deposit a layer of argon frost on the DIII-D neutral beam cryopanels, between tokamak injection pulses. The layer serves to trap helium on the cryopanels providing sufficient pumping speed for 5 s helium beam extraction. The argon frosting hardware is now present on two of four DIII-D neutral beamlines, allowing injection of up to 6 MW of helium neutral beams per discharge, with pulse lengths of up to 5 s. The argon frosting system is described, along with experimental results demonstrating its effectiveness as a method of economically extending the capabilities of cryogenic pumping panels to allow multi-second helium neutral beam injection

Hydrogen gas embrittlement of stainless steels mainly austenitic steels. Volumes 1 and 2

International Nuclear Information System (INIS)

Azou, P.

1988-01-01

Steel behavior in regard to hydrogen is examined especially austenitic steels. Gamma steels are studied particularly the series 300 with various stabilities and gamma steels with improved elasticity limit for intermetallic phase precipitation and nitrogen additions. A two-phase structure γ + α' is also studied. All the samples are tested for mechanical behavior in gaseous hydrogen. Influence of metallurgical effects and of testing conditions on hydrogen embrittlement are evidenced. Microstructure resulting from mechanical or heat treatments, dislocation motion during plastic deformation and influence of deformation rate are studied in detail [fr

Hydrogen embrittlement corrosion failure of water wall tubes in large power station boilers

International Nuclear Information System (INIS)

Mathur, P.K.

1981-01-01

In the present paper, causes and mechanism of hydrogen embrittlement failure of water wall tubes in high pressure boilers have been discussed. A low pH boiler water environment, produced as a result of condenser leakage or some other type of system contamination and presence of internal metal oxide deposits, which permit boiler water solids to concentrate during the process of steam generation, have been ascribed to accelerate the formation of local corrosion cells conducive for acid attack resulting in hydrogen damage failure of water wall tubes. (author)

An efficient continuous flow helium cooling unit for Moessbauer experiments

International Nuclear Information System (INIS)

Herbert, I.R.; Campbell, S.J.

1976-01-01

A Moessbauer continuous flow cooling unit for use with liquid helium over the temperature range 4.2 to 300K is described. The cooling unit can be used for either absorber or source studies in the horizontal plane and it is positioned directly on top of a helium storage vessel. The helium transfer line forms an integral part of the cooling unit and feeds directly into the storage vessel so that helium losses are kept to the minimum. The helium consumption is 0.12 l h -1 at 4.2 K decreasing to 0.055 l h -1 at 40 K. The unit is top loading and the exchange gas cooled samples can be changed easily and quickly. (author)

Helium implanted AlHf as studied by 181 Ta TDPAC

Indian Academy of Sciences (India)

Measurements on helium implanted sample indicate the binding of helium associated defects by Hf solute clusters. Isochronal annealing measurements indicate the dissociation of the helium implantation induced defects from Hf solute clusters for annealing treatments beyond 650 K. On comparison of the present results ...

The cryogenic helium cooling system for the Tokamak physics experiment

International Nuclear Information System (INIS)

Felker, B.; Slack, D.S.; Wendland, C.R.

1995-01-01

The Tokamak Physics Experiment (TPX) will use supercritical helium to cool all the magnets and supply helium to the Vacuum cryopumping subsystem. The heat loads will come from the standard steady state conduction and thermal radiation sources and from the pulsed loads of the nuclear and eddy currents caused by the Central Solenoid Coils and the plasma positioning coils. The operations of the TPX will begin with pulses of up to 1000 seconds in duration every 75 minutes. The helium system utilizes a pulse load leveling scheme to buffer out the effects of the pulse load and maintain a constant cryogenic plant operation. The pulse load leveling scheme utilizes the thermal mass of liquid and gaseous helium stored in a remote dewar to absorb the pulses of the tokamak loads. The mass of the stored helium will buffer out the temperature pulses allowing 5 K helium to be delivered to the magnets throughout the length of the pulse. The temperature of the dewar will remain below 5 K with all the energy of the pulse absorbed. This paper will present the details of the heat load sources, of the pulse load leveling scheme operations, a partial helium schematic, dewar temperature as a function of time, the heat load sources as a function of time and the helium temperature as a function of length along the various components that will be cooled

Film boiling heat transfer in liquid helium

International Nuclear Information System (INIS)

Inai, Nobuhiko

1979-01-01

The experimental data on the film boiling heat transfer in liquid helium are required for investigating the stability of superconducting wires. On the other hand, liquid helium has the extremely different physical properties as compared with the liquids at normal temperature such as water. In this study, the experiments on pool boiling were carried out, using the horizontal top surface of a 20 mm diameter copper cylinder in liquid helium. For observing individual bubbles, the experiments on film boiling from a horizontal platinum wire were performed separately in liquid nitrogen and liquid helium, and photographs of floating-away bubbles were taken. The author pointed out the considerable upward shift of the boiling curve near the least heat flux point in film boiling from the one given by the Berenson's equation which has been said to agree comparatively well with the data on the film boiling of the liquids at normal temperature, and the reason was investigated. Consequently, a model for film boiling heat transfer was presented. Also one equation expressing the film boiling at low heat flux for low temperature liquids was proposed. It represents well the tendency to shift from Berenson's equation of the experimental data on film boiling at the least heat flux point for liquid helium, liquid nitrogen and water having extremely different physical properties. Some discussions are added at the end of the paper. (Wakatsuki, Y.)

A description of stress driven bubble growth of helium implanted tungsten

International Nuclear Information System (INIS)

Sharafat, Shahram; Takahashi, Akiyuki; Nagasawa, Koji; Ghoniem, Nasr

2009-01-01

Low energy (<100 keV) helium implantation of tungsten has been shown to result in the formation of unusual surface morphologies over a large temperature range (700-2100 deg. C). Simulation of these macroscopic phenomena requires a multiscale approach to modeling helium transport in both space and time. We present here a multiscale helium transport model by coupling spatially-resolved kinetic rate theory (KRT) with kinetic Monte Carlo (KMC) simulation to model helium bubble nucleation and growth. The KRT-based HEROS Code establishes defect concentrations as well as stable helium bubble nuclei as a function of implantation parameters and position from the implanted surface and the KMC-based Mc-HEROS Code models the growth of helium bubbles due to migration and coalescence. Temperature- and stress-gradients can act as driving forces, resulting in biased bubble migration. The Mc-HEROS Code was modified to simulate the impact of stress gradients on bubble migration and coalescence. In this work, we report on bubble growth and gas release of helium implanted tungsten W/O stress gradients. First, surface pore densities and size distributions are compared with available experimental results for stress-free helium implantation conditions. Next, the impact of stress gradients on helium bubble evolution is simulated. The influence of stress fields on bubble and surface pore evolution are compared with stress-free simulations. It is shown that near surface stress gradients accelerate helium bubbles towards the free surface, but do not increasing average bubble diameters significantly.

Room temperature desorption of helium-3 from metal tritides

International Nuclear Information System (INIS)

Beavis, L.C.; Kass, W.J.

1976-10-01

It has long been known that helium-3 accumulates in metal tritides as tritium decays. Early in life nearly 100% of the helium-3 is retained in the lattice, but when a critical concentration is reached (material dependent), the lattice will no longer retain the helium-3 and it is emitted at about the generation rate. Measurements were recently made on a number of erbium tritides with varying concentrations in the ditritide phase. The expected early release characteristics are observed for all of the samples. However, ditritides with higher tritium concentrations reach the rapid release state at much lower helium-3 concentrations. For instance, the helium to metal concentration for rapid release in the unsaturated ditritide is about 0.22, whereas it is only one-tenth this value in the saturated ditritide. The additional tritium in the tritide appears to be the cause of this effect

Ultralow temperature helium compressor for Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Asakura, Hiroshi

1988-01-01

Ishikawajima Harima Heavy Industries Co., Ltd. started the development of an ultralow temperature helium compressor for helium liquefaction in 1984 jointly with Japan Atomic Energy Research Institute, and has delivered the first practical machine to the Superconductive Magnet Laboratory of JAERI. For a large superconductive magnet to be used in the stable state for a fusion reactor, conventional superconductive materials (NbTi, NbTi 3 Sn, etc.) must be used, being cooled forcibly with supercritical helium. The supercritical helium which is compressed above the critical pressure of 228 kPa has a stable cooling effect since the thermal conductivity does not change due to the evaporation of liquid helium. In order to maintain the temperature of the supercritical helium below 4 K before it enters a magnet, a heat exchanger is used. The compressor that IHI has developed has the ability to reduce the vapor pressure of liquid helium from atmospheric pressure to 50.7 kPa, and can attain the temperature of 3.5 K. The specification of this single stage centrifugal compressor is: mass flow rate 25 - 64 g/s, speed 80,000 rpm, adiabatic efficiency 62 - 69 %. The structure and the performance are reported. (K.I.)

Cooling performance of helium-gas/water coolers in HENDEL

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

1994-01-01

The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

Electron temperature measurements in lowdensity plasmas by helium spectroscopy

International Nuclear Information System (INIS)

Brenning, N.

1977-09-01

This method to use relative intensities of singlet and triplet lines of neutral helium to measure electron temperature in low-density plasmas is examined. Calculations from measured and theoretical data about transitions in neutral helium are carried out and compared to experimental results. It is found that relative intensities of singlet and triplet lines from neutral helium only can be used for TE determination in low-density, short-duration plasmas. The most important limiting processes are excitation from the metastable 2 3 S level and excitation transfer in collisions between electrons and excited helium atoms. An evaluation method is suggested, which minimizes the effect of these processes. (author)

Effect of the 718 alloy metallurgical status on hydrogen embrittlement; Effet de l'etat metallurgique de l'alliage 718 sur la fragilisation par l'hydrogene

Energy Technology Data Exchange (ETDEWEB)

Galvano, F.; Andrieu, E.; Blanc, Ch.; Odemer, G.; Ter-Ovanessian, B.; Cocheteau, N.; Holstein, A.; Reboul, Ch. [Universite de Toulouse, CIRIMAT, UPS/CNRS/INPT, 31 - Toulouse (France); Clouez, J.M. [AREVA NP 69 - Lyon (France)

2010-03-15

The Inconel 718 is a nickel superalloy which is widely used in the nuclear industry, but is sensitive to hydrogen embrittlement induced by corrosion and stress corrosion cracking phenomena, and by the presence of dissolved hydrogen in pressurized water reactor environments. As this alloy is hardened by precipitation of different intermetallic phases, it appeared that the presence of these precipitates has a strong influence on the hydrogen embrittlement. The authors report the study of the nature and effect of the different traps (intermetallic phases, carbides or their interfaces) on the hydrogen embrittlement susceptibility of the 718 alloy, and more particularly on the observed failure modes. Experiments are performed on tensile samples in which hydrogen content can be measured. The type and grain size of the observed microstructures are given with respect with the thermal treatment, as well as the mechanical properties with or without hydrogen loading

Thermal release behavior of helium from copper irradiated by He+ ions

International Nuclear Information System (INIS)

Yamauchi, T.; Tokura, S.; Yamanaka, S.; Miyake, M.

1988-01-01

Thermal release behavior of helium from copper irradiated by 20 keV He + ions with a dose of 2x10 15 to 3x10 17 ions/cm 2 has been studied. The shape of the thermal release curves and thew number of helium release peaks strongly depend on the irradiation dose. Results from SEM surface observastion after post-irradiation heating suggested that helium release caused various surface damages such as blistering, flaking, and hole formation. Helium release resulting in small holes was analyzed and helium bubble growth mechanisms are discussed. (orig.)

Radioactive ions and atoms in superfluid helium

NARCIS (Netherlands)

Dendooven, P.G.; Purushothaman, S.; Gloos, K.; Aysto, J.; Takahashi, N.; Huang, W.; Harissopulos, S; Demetriou, P; Julin, R

2006-01-01

We are investigating the use of superfluid helium as a medium to handle and manipulate radioactive ions and atoms. Preliminary results on the extraction of positive ions from superfluid helium at temperatures close to 1 K are described. Increasing the electric field up to 1.2 kV/cm did not improve

Conceptual design of helium experimental loop

International Nuclear Information System (INIS)

Yu Xingfu; Feng Kaiming

2007-01-01

In a future demonstration fusion power station (DEMO), helium is envisaged as coolant for plasma facing components, such as blanket and dive,or. All these components have a very complex geometry, with many parallel cooling channels, involving a complex helium flow distribution. Test blanket modules (TBM) of this concept will under go various tests in the experimental reactor ITER. For the qualification of TBM, it is indispensable to test mock-ups in a helium loop under realistic pressure and temperature profiles, in order to validate design codes, especially regarding mass flow and heat transition processes in narrow cooling channels. Similar testing must be performed for DEMO blanket, currently under development. A Helium Experimental Loop (HELOOP) is planed to be built for TBM tests. The design parameter of temperature, pressure, flow rate is 550 degree C, 10 MPa, l kg/s respectively. In particular, HELOOP is able to: perform full-scale tests of TBM under realistic conditions; test other components of the He-cooling system in ITER; qualify the purification circuit; obtain information for the design of the ITER cooling system. The main requirements and characteristics of the HELOOP facility and a preliminary conceptual design are described in the paper. (authors)

Resource letter SH-1: superfluid helium

International Nuclear Information System (INIS)

Hallock, R.B.

1982-01-01

The resource letter covers the general subject of superfluid helium and treats 3 He and 3 He-- 4 He mixtures as well as 4 He. No effort has been made to include the fascinating experiments on either solid helium or the equally fascinating work on adsorbed helium where the helium coverage is below that necessary for superfluidity. An earlier resource letter by C. T. Lane [Am. J. Phys. 35, 367 (1967)] may be consulted for additional comments on some of the cited earlier manuscripts, but the present work is self-contained and may be used independently. Many high-quality research reports have not been cited here. Rather, the author has tried in most cases to include works particularly readable or relevant. There is a relatively heavy emphasis on experimental references. The primary reason is that these works tend to be more generally readable. No doubt some works that might have been included, have not, and for this the author takes responsibility with apology. Articles selected for incorporation in a reprint volume (to be published separately by the American Association of Physics Teachers) are marked with an asterisk(*). Following each referenced work the general level of difficulty is indicated by E, I, or A for elementary, intermediate, or advanced

Progress in identification of radiation embrittlement mechanisms

International Nuclear Information System (INIS)

Hawthorne, J.R.

1988-01-01

This report outlines recent advances in the isolation and understanding of mechanisms behind known composition influences on he radiation embrittlement sensitivity of reactor pressure vessel steels at 288 deg. C. The advances are largely the product of joint investigations by Materials Engineering Associates (MEA) and other laboratories in the U.S. and overseas under cooperative and subcontract arrangements. Specific objectives were: confirmation of the suspect Cu mechanism, identification of the process for the Cu:Ni synergism, and isolation of the P mechanism in radiation sensitivity development. The investigations proceeded with MEA-supplied steels and iron alloys from 4-way split laboratory melts; research tools included Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM), Field Ion Microscopy (FIM), Small Angle Neutron Scattering (SANS), Positron Annihilation (PA) and Auger Electron Spectroscopy (AES). Experimental results show that P and Cu enhance the radiation elevation of yield strength and that the associated mechanisms are a radiation-induced precipitation of P or Cu-rich clusters which impede dislocation motion. With high Cu alloys, a Cu phosphide is formed in preference to P precipitates and the P contribution is greatly reduced. Effects of postirradiation annealing and reirradiation are also reported. (author)

High-efficiency pump for space helium transfer. Final Technical Report

International Nuclear Information System (INIS)

Hasenbein, R.; Izenson, M.G.; Swift, W.L.; Sixsmith, H.

1991-12-01

A centrifugal pump was developed for the efficient and reliable transfer of liquid helium in space. The pump can be used to refill cryostats on orbiting satellites which use liquid helium for refrigeration at extremely low temperatures. The pump meets the head and flow requirements of on-orbit helium transfer: a flow rate of 800 L/hr at a head of 128 J/kg. The overall pump efficiency at the design point is 0.45. The design head and flow requirements are met with zero net positive suction head, which is the condition in an orbiting helium supply Dewar. The mass transfer efficiency calculated for a space transfer operation is 0.99. Steel ball bearings are used with gas fiber-reinforced teflon retainers to provide solid lubrication. These bearings have demonstrated the longest life in liquid helium endurance tests under simulated pumping conditions. Technology developed in the project also has application for liquid helium circulation in terrestrial facilities and for transfer of cryogenic rocket propellants in space

Fission neutron irradiation of copper containing implanted and transmutation produced helium

DEFF Research Database (Denmark)

Singh, B.N.; Horsewell, A.; Eldrup, Morten Mostgaard

1992-01-01

High purity copper containing approximately 100 appm helium was produced in two ways. In the first, helium was implanted by cyclotron at Harwell at 323 K. In the second method, helium was produced as a transmutation product in 800 MeV proton irradiation at Los Alamos, also at 323 K. The distribut......High purity copper containing approximately 100 appm helium was produced in two ways. In the first, helium was implanted by cyclotron at Harwell at 323 K. In the second method, helium was produced as a transmutation product in 800 MeV proton irradiation at Los Alamos, also at 323 K...... as well as the effect of the presence of other transmutation produced impurity atoms in the 800 MeV proton irradiated copper will be discussed....

Tables of thermodynamic properties of helium magnet coolant

International Nuclear Information System (INIS)

McAshan, M.

1992-07-01

The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: ''Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10 8 Pa'', Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923--1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: ''Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres'', Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it

Helium transport and exhaust studies in enhanced confinement regimes in DIII-D

International Nuclear Information System (INIS)

Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Burrell, K.H.; Finkenthal, D.F.; Gohil, P.; Groebner, R.J.

1995-02-01

A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust (τ* He /τ E ∼ 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D He /X eff ∼1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system

Temper embrittlement, irradiation induced phosphorus segregation and implications for post-irradiation annealing of reactor pressure vessels

International Nuclear Information System (INIS)

McElroy, R.J.; English, C.A.; Foreman, A.J.; Gage, G.; Hyde, J.M.; Ray, P.H.N.; Vatter, I.A.

1999-01-01

Three steels designated JPB, JPC and JPG from the IAEA Phase 3 Programme containing two copper and phosphorus levels were pre- and post-irradiation Charpy and hardness tested in the as-received (AR), 1200 C/0.5h heat treated (HT) and heat treated and 450 C/2000h aged (HTA) conditions. The HT condition was designed to simulate coarse grained heat-affected zones (HAZ's) and showed a marked sensitivity to thermal ageing in all three alloys. Embrittlement after thermal ageing was greater in the higher phosphorus alloys JPB and JPG. Charpy shifts due to thermal ageing of between 118 and 209 C were observed and accompanied by pronounced intergranular fracture, due to phosphorus segregation. The irradiation embrittlement response was complex. The low copper alloys, JPC and JPB, in the HT and HTA condition exhibited significant irradiation induced Charpy shift but very low or even negative hardness changes indicating non-hardening embrittlement. The higher copper alloy, JPG, also exhibited irradiation hardening in line with its copper content. Fractographic and microchemical studies indicated irradiation induced phosphorus segregation and a transition from cleavage to intergranular failure at grain boundary phosphorus concentrations above a critical level. The enhanced grain boundary phosphorus level increased with dose in agreement with a kinetic segregation model developed at Harwell. The relevance of the thermal ageing studies to RPV Annealing for Plant-Life Extension was identified early in the program. It is of concern that annealing of RPV's has been performed, or is proposed, at temperatures in the range 425--475 C for periods of about 1 week (168h). Much attention has been given to the use of in-situ hardness measurements and machining miniature Charpy and tensile specimens from belt-line plate and weld materials. However, HAZ's, often containing higher phosphorus levels than the present materials, have largely been ignored. A post-irradiation annealing (PIA

THE POPULATION OF HELIUM-MERGER PROGENITORS: OBSERVATIONAL PREDICTIONS

International Nuclear Information System (INIS)

Fryer, Chris L.; Belczynski, Krzysztof; Bulik, Tomasz; Berger, Edo; Thöne, Christina; Ellinger, Carola

2013-01-01

The helium-merger gamma-ray burst (GRB) progenitor is produced by the rapid accretion onto a compact remnant (neutron star or black hole) when it undergoes a common envelope inspiral with its companion's helium core. This merger phase produces a very distinct environment around these outbursts and recent observations suggest that, in some cases, we are detecting the signatures of the past merger in the GRB afterglow. These observations allow us, for the first time, to study the specific features of the helium-merger progenitor. In this paper, we couple population synthesis calculations to our current understanding of GRB engines and common envelope evolution to make observational predictions for the helium-merger GRB population. Many mergers do not produce GRB outbursts and we discuss the implications of these mergers with the broader population of astrophysical transients.

Tensile properties of ADI material in water and gaseous environments

Energy Technology Data Exchange (ETDEWEB)

Rajnovic, Dragan, E-mail: draganr@uns.ac.rs [Faculty of Technical Sciences, University of Novi Sad, Trg D. Obradovića 6, 21000 Novi Sad (Serbia); Balos, Sebastian; Sidjanin, Leposava [Faculty of Technical Sciences, University of Novi Sad, Trg D. Obradovića 6, 21000 Novi Sad (Serbia); Eric Cekic, Olivera [Innovation Centre, Faculty of Mechanical Engineering, University of Belgrade, Kraljice Marije 16, 11120 Belgrade (Serbia); Grbovic Novakovic, Jasmina [Vinca Institute of Nuclear Sciences, University of Belgrade, P.O. Box 522, 11000 Belgrade (Serbia)

2015-03-15

Austempered ductile iron (ADI) is an advanced type of heat treated ductile iron, having comparable mechanical properties as forged steels. However, it was found that in contact with water the mechanical properties of austempered ductile irons decrease, especially their ductility. Despite considerable scientific attention, the cause of this phenomenon remains unclear. Some authors suggested that hydrogen or small atom chemisorption causes the weakening of the surface atomic bonds. To get additional reliable data of that phenomenon, in this paper, two different types of austempered ductile irons were tensile tested in various environments, such as: argon, helium, hydrogen gas and water. It was found that only the hydrogen gas and water gave a statistically significant decrease in mechanical properties, i.e. cause embrittlement. Furthermore, the fracture surface analysis revealed that the morphology of the embrittled zone near the specimen surface shares similarities to the fatigue micro-containing striation-like lines, which indicates that the morphology of the brittle zone may be caused by cyclic local-chemisorption, micro-embrittlement and local-fracture. - Highlights: • In contact with water and other liquids the ADI suddenly exhibits embrittlement. • The embrittlement is more pronounced in water than in the gaseous hydrogen. • The hydrogen chemisorption into ADI surface causes the formation of a brittle zone. • The ADI austempered at lower temperatures (300 °C) is more resistant to embrittlement.

New technique for enhancing helium production in ferritic materials

International Nuclear Information System (INIS)

Greenwood, L.R.; Graczyk, D.G.; Kneff, D.W.

1987-10-01

Analyses of iron samples irradiated up to 10 27 n/m 2 in HFIR found more helium than was expected from fast neutron reactions at high neutron fluences. The helium excess increases systematically with neutron exposure, suggesting a transmutation-driven process. The extra helium may be produced in two different ways, either by fast neutron reactions on the transmuted isotopes of iron or by a thermal neutron reaction with the radioactive isotope 55 Fe. Radiometric and mass spectrometric measurements of the iron isotopes composing the irradiated samples have been used to determine limits on the cross sections for each process. Either of these processes can be used to enhance helium production in ferritic materials during irradiations in mixed-spectrum reactors by isotopically enriching the samples. Further work is needed to clarify the reaction mechanisms and helium production cross sections. Our measurements determined the thermal neutron total absorption cross section of 55 Fe to be 13.2 +- 2.1 barns. 16 refs., 3 figs., 3 tabs

Helium Leak Test for the PLS Storage Ring Chamber

International Nuclear Information System (INIS)

Choi, M. H.; Kim, H. J.; Choi, W. C.

1993-01-01

The storage ring vacuum system for the Pohang Light Source (PLS) has been designed to maintain the vacuum pressure of 10 1 0 Torr which requires UHV welding to have helium leak rate less than 1x10 1 0 Torr·L/sec. In order to develop new technique (PLS) welding technique), a prototype vacuum chamber has been welded by using Tungsten Inert Gas welding method and all the welded joints have been tested with a non-destructive method, so called helium leak detection, to investigate the vacuum tightness of the weld joints. The test was performed with a detection limit of 1x10 1 0 Torr·L/sec for helium and no detectable leaks were found for all the welded joints. Thus the performance of welding technique is proven to meet the criteria of helium leak rate required in the PLS Storage Ring. Both the principle and the procedure for the helium leak detection are also discussed

Helium bubble formation and retention in Cu-Nb nanocomposites

International Nuclear Information System (INIS)

Dunn, A.Y.; McPhie, M.G.; Capolungo, L.; Martinez, E; Cherkaoui, M.

2013-01-01

A spatially dependent rate theory model for helium migration, clustering, and trapping on interfaces between Cu and Nb layers is introduced to predict the evolution of the concentrations of He clusters of various sizes during implantation and early annealing. Migration and binding energies of point defects and small clusters in bulk Cu and Nb are found using conjugate gradient minimization and the nudged elastic band method. The model is implemented in a three-dimensional framework and used to predict the relationship between helium bubble formation and the nano-composite microstructure, including interfacial free volume, grain size, and layer thickness. Interstitial and vacancy-like migration of helium is considered. The effects of changing layer thickness and interfacial misfit dislocation density on the threshold for helium bubble nucleation are found to match experiments. Accelerated helium release due to interfaces and grain boundaries is shown to occur only when diffusion rates on interfaces and grain boundaries are greatly increased relative to the bulk material.

Biomolecular ions in superfluid helium nanodroplets

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Florez, Ana Isabel

2016-07-01

The function of a biological molecule is closely related to its structure. As a result, understanding and predicting biomolecular structure has become the focus of an extensive field of research. However, the investigation of molecular structure can be hampered by two main difficulties: the inherent complications that may arise from studying biological molecules in their native environment, and the potential congestion of the experimental results as a consequence of the large number of degrees of freedom present in these molecules. In this work, a new experimental setup has been developed and established in order to overcome the afore mentioned limitations combining structure-sensitive gas-phase methods with superfluid helium droplets. First, biological molecules are ionised and brought into the gas phase, often referred to as a clean-room environment, where the species of interest are isolated from their surroundings and, thus, intermolecular interactions are absent. The mass-to-charge selected biomolecules are then embedded inside clusters of superfluid helium with an equilibrium temperature of ∝0.37 K. As a result, the internal energy of the molecules is lowered, thereby reducing the number of populated quantum states. Finally, the local hydrogen bonding patterns of the molecules are investigated by probing specific vibrational modes using the Fritz Haber Institute's free electron laser as a source of infrared radiation. Although the structure of a wide variety of molecules has been studied making use of the sub-Kelvin environment provided by superfluid helium droplets, the suitability of this method for the investigation of biological molecular ions was still unclear. However, the experimental results presented in this thesis demonstrate the applicability of this experimental approach in order to study the structure of intact, large biomolecular ions and the first vibrational spectrum of the protonated pentapeptide leu-enkephalin embedded in helium

Biomolecular ions in superfluid helium nanodroplets

International Nuclear Information System (INIS)

Gonzalez Florez, Ana Isabel

2016-01-01

The function of a biological molecule is closely related to its structure. As a result, understanding and predicting biomolecular structure has become the focus of an extensive field of research. However, the investigation of molecular structure can be hampered by two main difficulties: the inherent complications that may arise from studying biological molecules in their native environment, and the potential congestion of the experimental results as a consequence of the large number of degrees of freedom present in these molecules. In this work, a new experimental setup has been developed and established in order to overcome the afore mentioned limitations combining structure-sensitive gas-phase methods with superfluid helium droplets. First, biological molecules are ionised and brought into the gas phase, often referred to as a clean-room environment, where the species of interest are isolated from their surroundings and, thus, intermolecular interactions are absent. The mass-to-charge selected biomolecules are then embedded inside clusters of superfluid helium with an equilibrium temperature of ∝0.37 K. As a result, the internal energy of the molecules is lowered, thereby reducing the number of populated quantum states. Finally, the local hydrogen bonding patterns of the molecules are investigated by probing specific vibrational modes using the Fritz Haber Institute's free electron laser as a source of infrared radiation. Although the structure of a wide variety of molecules has been studied making use of the sub-Kelvin environment provided by superfluid helium droplets, the suitability of this method for the investigation of biological molecular ions was still unclear. However, the experimental results presented in this thesis demonstrate the applicability of this experimental approach in order to study the structure of intact, large biomolecular ions and the first vibrational spectrum of the protonated pentapeptide leu-enkephalin embedded in helium