Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite; Realisation d'un appareil de mesure de la conductibilite thermique du graphite irradie ou non irradie

Energy Technology Data Exchange (ETDEWEB)

Bocquet, M; Micaud, G

1962-07-01

An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient {delta}{theta}/ {delta}x in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. {delta}x/{delta}{theta}, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [French] Un appareil de mesure du coefficient de conductibilite thermique K du graphite irradie ou non irradie a ete realise. Utilisant le principe du transfert de chaleur, il permet de mesurer K au voisinage de la temperature ambiante avec une precision de 6 pour cent environ. L'echantillon de graphite etudie est place sous vide entre une source chaude et une source froide qui creent en regime permanent un gradient de temperature {delta}{theta}/{delta}x La quantite de chaleur transferee Q est deduite de la puissance electrique dissipee dans la source chaude en deduisant les pertes thermiques. Le coefficient de conductibilite thermique est defini par: K = Q/S. {delta}x/{delta}{theta} S designant la section de l'echantillon. Des etudes systematiques ont permis de determiner pour differents graphites non irradies les valeurs moyennes des coefficients de conductibilite thermique. Ces etudes ont mis en evidence pour un type de graphite donne, l'influence de la densite apparente sur le coefficient de conductibilite thermique. A partir de mesures effectuees sur des echantillons de graphite irradies preleves par carottage dans les empilements des reacteurs a moderateur de graphite les variations du rapport K0/Ki en fonction de la dose et de la

Significance of primary irradiation creep in graphite

CSIR Research Space (South Africa)

Erasmus, C

2013-05-01

Full Text Available Traditionally primary irradiation creep is introduced into graphite analysis by applying the appropriate amount of creep strain to the model at the initial time-step. This is valid for graphite components that are subjected to high fast neutron flux...

The effect of compressive stress on the Young's modulus of unirradiated and irradiated nuclear graphites

International Nuclear Information System (INIS)

Oku, T.; Usui, T.; Ero, M.; Fukuda, Y.

1977-01-01

The Young's moduli of unirradiated and high temperature (800 to 1000 0 C) irradiated graphites for HTGR were measured by the ultrasonic method in the direction of applied compressive stress during and after stressing. The Young's moduli of all the tested graphites decreased with increasing compressive stress both during and after stressing. In order to investigate the reason for the decrease in Young's modulus by applying compressive stress, the mercury pore diameter distributions of a part of the unirradiated and irradiated specimens were measured. The change in pore distribution is believed to be associated with structural changes produced by irradiation and compressive stressing. The residual strain, after removing the compressive stress, showed a good correlation with the decrease in Young's modulus caused by the compressive stress. The decrease in Young's modulus by applying compressive stress was considered to be due to the increase in the mobile dislocation density and the growth or formation of cracks. The results suggest, however, that the mechanism giving the larger contribution depends on the brand of graphite, and in anisotropic graphite it depends on the direction of applied stress and the irradiation conditions. (author)

Isotropic nuclear graphites; the effect of neutron irradiation

International Nuclear Information System (INIS)

Lore, J.; Buscaillon, A.; Mottet, P.; Micaud, G.

1977-01-01

Several isotropic graphites have been manufactured using different forming processes and fillers such as needle coke, regular coke, or pitch coke. Their properties are described in this paper. Specimens of these products have been irradiated in the fast reactor Rapsodie between 400 to 1400 0 C, at fluences up to 1,7.10 21 n.cm -2 PHI.FG. The results show an isotropic behavior under neutron irradiation, but the induced dimensional changes are higher than those of isotropic coke graphites although they are lower than those of conventional extruded graphites made with the same coke

Property changes in graphite irradiated at changing irradiation temperature

International Nuclear Information System (INIS)

Price, R.J.; Haag, G.

1979-07-01

Design data for irradiated graphite are usually presented as families of isothermal curves showing the change in physical property as a function of fast neutron fluence. In this report, procedures for combining isothermal curves to predict behavior under changing irradiation temperatures are compared with experimental data on irradiation-induced changes in dimensions, Young's modulus, thermal conductivity, and thermal expansivity. The suggested procedure fits the data quite well and is physically realistic

Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite: consequences on the mobility of chlorine 36 in irradiated graphites

International Nuclear Information System (INIS)

Blondel, Antoine

2013-01-01

This thesis deals with the studies of the management of irradiated graphite wastes issued from the dismantling of the UNGG French reactors. This work focuses on the behavior of 36 Cl. This radionuclide is mainly issued through the neutron activation of 35 Cl by the reaction 35 Cl(n, γ) 36 Cl, pristine chlorine being an impurity of nuclear graphite, present at the level of some at.ppm. 36 Cl is a long lived radionuclide (about 300,000 years) and is highly soluble in water and mobile in concrete and clay. The solubilization of 36 Cl is controlled by the water accessibility into irradiated graphite pores as well as by factors related to 36 Cl itself such as its chemical speciation and its location within the irradiated graphite. Both speciation and chlorine location should strongly influence its behaviour and need to be taken into account for the choice of liable management options. However, data on radioactive chlorine features are difficult to assess in irradiated graphite and are mainly related to detection sensitivity problems. In this context, we simulated and evaluated the impact of the temperature, the irradiation and the radiolytic oxidation on the chlorine 36 behaviour. In order to simulate the presence of 36 Cl, we implanted 37 Cl into virgin nuclear graphite. Ion implantation has been widely used to study the lattice location, the diffusion and the release of fission and activation products in nuclear materials. Our results on the comparative effects of the temperature and the irradiation show that chlorine occurs in irradiated graphite on temperature and electronic and nuclear irradiation improve this effect. (author)

Some physical methods for study of irradiation effects in graphite; Quelques procedes physiques pour etudier les effets de l'irradiation du graphite

Energy Technology Data Exchange (ETDEWEB)

Mayer, G; Lecomte, M; Mattmuller, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

A calibration method for a classical apparatus for differential thermal analysis is described in detail. This method achieves a relative precision of 5 per cent in the measurement of the internal energy release accompanying the annealing of irradiated graphites. Elastic constants of graphites are obtained from the frequencies of the longitudinal modes of vibration; procedures for excitation and detection of these vibrations at any temperature between -190 deg. C and +1500 deg. C are described. A procedure for obtaining easily measured deformations of graphites after relatively little irradiation with thermal neutrons is discussed. An application of this method to the study of the thermal annealing of elongation caused by displaced atoms is indicated. (author) [French] On decrit en detail une methode d'etalonnage pour un appareil classique d'analyse thermique differentielle. Cette methode permet de mesurer avec une precision relative de 5% la liberation d'energie interne qui accompagne le 'recuit' des graphites irradies. On deduit les constantes elastiques des graphites des frequences des vibrations longitudinales et on decrit les procedes pour exciter et detecter ces vibrations a toutes les temperatures comprises entre -190 deg. C et + 1500 deg. C. On discute un procede pour obtenir une des deformations de graphites facilement mesurables apres une irradiation relativement faible a l'aide de neutrons thermiques. Une application de cette methode a l'etude du 'recuit' thermique de l'elongation causee par les atomes deplaces est indiquee. (auteur)

High-temperature irradiation effects on mechnical properties of HTGR graphites

International Nuclear Information System (INIS)

Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

1978-04-01

The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

Examination of Experimental Data for Irradiation - Creep in Nuclear Graphite

Science.gov (United States)

Mobasheran, Amir Sassan

The objective of this dissertation was to establish credibility and confidence levels of the observed behavior of nuclear graphite in neutron irradiation environment. Available experimental data associated with the OC-series irradiation -induced creep experiments performed at the Oak Ridge National Laboratory (ORNL) were examined. Pre- and postirradiation measurement data were studied considering "linear" and "nonlinear" creep models. The nonlinear creep model considers the creep coefficient to vary with neutron fluence due to the densification of graphite with neutron irradiation. Within the range of neutron fluence involved (up to 0.53 times 10^{26} neutrons/m ^2, E > 50 KeV), both models were capable of explaining about 96% and 80% of the variation of the irradiation-induced creep strain with neutron fluence at temperatures of 600^circC and 900^circC, respectively. Temperature and reactor power data were analyzed to determine the best estimates for the actual irradiation temperatures. It was determined according to thermocouple readouts that the best estimate values for the irradiation temperatures were well within +/-10 ^circC of the design temperatures of 600^circC and 900 ^circC. The dependence of the secondary creep coefficients (for both linear and nonlinear models) on irradiation temperature was determined assuming that the variation of creep coefficient with temperature, in the temperature range studied, is reasonably linear. It was concluded that the variability in estimate of the creep coefficients is definitely not the results of temperature fluctuations in the experiment. The coefficients for the constitutive equation describing the overall growth of grade H-451 graphite were also studied. It was revealed that the modulus of elasticity and the shear modulus are not affected by creep and that the electrical resistivity is slightly (less than 5%) changed by creep. However, the coefficient of thermal expansion does change with creep. The consistency of

Proton irradiated graphite grades for a long baseline neutrino facility experiment

International Nuclear Information System (INIS)

Simos, N.; Nocera, P.; Zwaska, R.; Mokhov, N.

2017-01-01

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10"2"0 p/cm"2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10"2"0 cm"-"2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

Proton irradiated graphite grades for a long baseline neutrino facility experiment

Directory of Open Access Journals (Sweden)

N. Simos

2017-07-01

Full Text Available In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF of the Deep Underground Neutrino Experiment (DUNE four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ∼6.1×10^{20}  p/cm^{2} and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ∼10^{20}  cm^{−2} where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite

Proton irradiated graphite grades for a long baseline neutrino facility experiment

Science.gov (United States)

Simos, N.; Nocera, P.; Zhong, Z.; Zwaska, R.; Mokhov, N.; Misek, J.; Ammigan, K.; Hurh, P.; Kotsina, Z.

2017-07-01

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140-180 MeV, to peak fluence of ˜6.1 ×1020 p /cm2 and irradiation temperatures between 120 - 200 °C . The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of ˜1020 cm-2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

Project accent: graphite irradiated creep in a materials test reactor

International Nuclear Information System (INIS)

Brooking, M.

2014-01-01

Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

2014-01-01

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

International Nuclear Information System (INIS)

Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

2001-01-01

It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

International Nuclear Information System (INIS)

Anfimov, S.S.

2001-01-01

In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination

International Nuclear Information System (INIS)

Brown, F.J.; Palmer, J.D.; Wood, P.

2001-01-01

The irradiation of materials in nuclear reactors results in neutron activation of component elements. Irradiated graphite wastes arise from their use in UK gas-cooled research and commercial reactor cores, and in fuel element components, where the graphite has acted as the neutron moderator. During irradiation the residual chlorine, which was used to purify the graphite during manufacture, is activated to chlorine-36. This isotope is long-lived and poorly retarded by geological barriers, and may therefore be a key radionuclide with respect to post-closure disposal facilities performance. United Kingdom Nirex Limited, currently responsible for the development of a disposal route for intermediate-level radioactive wastes in the UK, carried out a major research programme to support an overall assessment of the chlorine-36 activity of all wastes including graphite reactor components. The various UK gas cooled reactors reactors have used a range of graphite components made from diverse graphite types; this has necessitated a systematic programme to cover the wide range of graphite and production processes. The programme consisted of: precursor measurements - on the surface and/or bulk of representative samples of relevant materials, using specially developed methods; transfer studies - to quantify the potential for transfer of Cl-36 into and between waste streams during irradiation of graphite; theoretical assessments - to support the calculational methodology; actual measurements - to confirm the modelling. For graphite, a total of 458 measurements on samples from 57 batches were performed, to provide a detailed understanding of the composition of nuclear graphite. The work has resulted in the generation of probability density functions (PDF) for the mean chlorine concentration of three classes of graphite: fuel element graphite; Magnox moderator and reflector graphite and AGR reflector graphite; AGR moderator graphite. Transfer studies have shown that a significant

Variation of the properties of siliconized graphite during neutron irradiation

International Nuclear Information System (INIS)

Virgil'ev, Y.S.; Chugunova, T.K.; Pikulik, R.G.

1986-01-01

The authors evaluate the radiation-induced property changes in siliconized graphite of the industrial grades SG-P and SG-M. The authors simultaneously tested the reference (control) specimens of graphite that are used as the base for obtaining the SG-M siliconized graphite by impregnating with silicon. The suggested scheme (model) atributes the dimensional changes of the siliconized graphite specimens to the effect of the quantitative ratio of the carbide phase and carbon under different conditions of irradiation. If silicon is insufficient for the formation of a dense skeleton, graphite plays a devisive role, and it may be assumed that at an irradiation temperature greater than 600 K, the material shrinks. The presence of isolated carbide inclusions also affects the physicomechanical properties (including the anitfriction properties)

Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

International Nuclear Information System (INIS)

Marcu, A.; Avotina, L.; Porosnicu, C.; Marin, A.; Grigorescu, C.E.A.; Ursescu, D.; Lungu, M.; Demitri, N.; Lungu, C.P.

2015-01-01

Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp 3 % increase at tens of nm below the surface. • sp 3 % is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp 3 bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp 3 percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

Temperature annealing of tracks induced by ion irradiation of graphite

International Nuclear Information System (INIS)

Liu, J.; Yao, H.J.; Sun, Y.M.; Duan, J.L.; Hou, M.D.; Mo, D.; Wang, Z.G.; Jin, Y.F.; Abe, H.; Li, Z.C.; Sekimura, N.

2006-01-01

Highly oriented pyrolytic graphite (HOPG) samples were irradiated by Xe ions of initial kinetic energy of 3 MeV/u. The irradiations were performed at temperatures of 500 and 800 K. Scanning tunneling microscopy (STM) images show that the tracks occasionally have elongated structures under high-temperature irradiation. The track creation yield at 800 K is by three orders of magnitude smaller compared to that obtained during room-temperature irradiation. STM and Raman spectra show that amorphization occurs in graphite samples irradiated at 500 K to higher fluences, but not at 800 K. The obtained experimental results clearly reveal that the irradiation under high temperature causes track annealing

Erosion of pyrolytic graphite and Ti-doped graphite due to high flux irradiation

International Nuclear Information System (INIS)

Ohtsuka, Yusuke; Ohashi, Junpei; Ueda, Yoshio; Isobe, Michiro; Nishikawa, Masahiro

1997-01-01

The erosion of pyrolytic graphite and titanium doped graphite RG-Ti above 1,780 K was investigated by 5 keV Ar beam irradiation with the flux from 4x10 19 to 1x10 21 m -2 ·s -1 . The total erosion yields were significantly reduced with the flux. This reduction would be attributed to the reduction of RES (radiation enhanced sublimation) yield, which was observed in the case of isotropic graphite with the flux dependence of RES yield of φ -0.26 (φ: flux) obtained in our previous work. The yield of pyrolytic graphite was roughly 30% higher than that of isotropic graphite below the flux of 10 20 m -2 ·s -1 whereas each yield approached to very close value at the highest flux of 1x10 21 m -2 ·s -1 . This result indicated that the effect of graphite structure on the RES yield, which was apparent in the low flux region, would disappear in the high flux region probably due to the disordering of crystal structure. In the case of irradiation to RG-Ti at 1,780 K, the surface undulations evolved with a mean height of about 3 μm at 1.2x10 20 m -2 ·s -1 , while at higher flux of 8.0x10 20 m -2 ·s -1 they were unrecognizable. These phenomena can be explained by the reduction of RES of graphite parts excluding TiC grains. (author)

Electron spin resonance in neutron-irradiated graphite. Dependence on temperature and effect of annealing; Resonance paramagnetique du graphite irradie aux neutrons. Variation en fonction de la temperature et experiences de recuit

Energy Technology Data Exchange (ETDEWEB)

Kester, T [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires, Laboratoire de resonance magnetique

1967-09-01

The temperature dependence of the electron spin resonance signal from neutron irradiated graphite has been studied. The results lead to an interpretation of the nature of the paramagnetic centers created by irradiation. In annealing experiments on graphite samples, which had been irradiated at low temperature, two annealing peaks and one anti-annealing peak were found. Interpretations are proposed for these peaks. (author) [French] Le graphite irradie aux neutrons a ete etudie par resonance paramagnetique electronique en fonction de la temperature. La nature des centres paramagnetiques crees par irradiation est interpretee a l'aide des resultats. Des experiences de recuit sur des echantillons de graphite irradie a 77 deg. K ont permis de mettre en evidence deux pics de recuit et un pic d'anti-recuit, pour lesquels des interpretations sont proposees. (auteur)

Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

Energy Technology Data Exchange (ETDEWEB)

Marcu, A., E-mail: aurelian.marcu@inflpr.ro [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Avotina, L. [Institute of Chemical Physics, University of Latvia, Kronvalda 4, LV 1010 Riga (Latvia); Porosnicu, C. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Marin, A. [Ilie Murgulescu” Institute of Physical Chemistry, 202 Splaiul Independentei 060021, Bucharest (Romania); Grigorescu, C.E.A. [National Institute R& D for Optoelectronics INOE 2000, 077125 Bucharest (Romania); Ursescu, D. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Lungu, M. [National Institute of Materials Physics Atomistilor Str., 105 bis, 077125, Magurele (Romania); Demitri, N. [Hard X-ray Beamline and Structural Biology, Elettra-Sincrotrone Trieste, Strada Statale 14 - km 163,5 in AREA Science Park, 34149 Basovizza TS Italy (Italy); Lungu, C.P. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania)

2015-11-15

Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp{sup 3}% increase at tens of nm below the surface. • sp{sup 3}% is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp{sup 3} bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp{sup 3} percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

Transmission electron-microscopic studies of structural changes in polycrystalline graphite after high temperature irradiation

International Nuclear Information System (INIS)

Platonov, P.A.; Gurovich, B.A.; Shtrombakh, Ya.I.; Karpukhin, V.I.

1985-01-01

Transmission electron-microscopic investigation of polycrystalline graphite before and after irradiation is carried out. The direct use of graphite samples after ion thinning, as an inquiry subject is the basic peculiarity of the work. Main structural components of MPG-6 graphite before and after irradiation are revealed, the structural mechanism of the reactor graphite destruction under irradiation is demonstrated. The mean values of L αm and L cm crystallite dimensions are determined. Radiation defects, occuring in some crystallites after irradiation are revealed by the dark-field electron microscopy method

Curling and closure of graphitic networks under electron-beam irradiation

Energy Technology Data Exchange (ETDEWEB)

Ugarte, D [Ecole Polytechnique Federale, Lausanne (Switzerland)

1992-10-22

The discovery of buckminsterfullerene (C[sub 60]) and its production in macroscopic quantities has stimulated a great deal of research. More recently, attention has turned towards other curved graphitic networks, such as the giant fullerenes (C[sub n], n > 100) and carbon nanotubes. A general mechanism has been proposed in which the graphitic sheets bend in an attempt to eliminate the highly energetic dangling bonds present at the edge of the growing structure. Here, I report the response of carbon soot particles and tubular graphitic structures to intense electron-beam irradiation in a high-resolution electron microscope; such conditions resemble a high-temperature regime, permitting a degree of structural fluidity. With increased irradiation, there is a gradual reorganization of the initial material into quasi-spherical particles composed of concentric graphitic shells. This lends weight to the nucleation scheme proposed for fullerenes, and moreover, suggests that planar graphite may not be the most stable allotrope of carbon in systems of limited size. (Author).

A 2-D nucleation-growth model of spheroidal graphite

International Nuclear Information System (INIS)

Lacaze, Jacques; Bourdie, Jacques; Castro-Román, Manuel Jesus

2017-01-01

Analysis of recent experimental investigations, in particular by transmission electron microscopy, suggests spheroidal graphite grows by 2-D nucleation of new graphite layers at the outer surface of the nodules. These layers spread over the surface along the prismatic direction of graphite which is the energetically preferred growth direction of graphite when the apparent growth direction of the nodules is along the basal direction of graphite. 2-D nucleation-growth models first developed for precipitation of pure substances are then adapted to graphite growth from the liquid in spheroidal graphite cast irons. Lateral extension of the new graphite layers is controlled by carbon diffusion in the liquid. This allows describing quantitatively previous experimental results giving strong support to this approach.

The utilization of a pressurized-graphite/water/oxygen mixture for irradiated graphite incineration

International Nuclear Information System (INIS)

Antonini, G.; Perotin, J.P.; Charlot, P.

1992-01-01

The authors demonstrate the interest of the utilization of a pressurized-graphite/water/oxygen mixture in the incineration of irradiated graphite. The aqueous phase comes in the form of a three-dimensional system that traps pressurized oxygen, the pulverulent solid being dispersed at the liquid/gas interfaces. These three-phasic formulations give the following advantages: reduction of the apparent viscosity of the mixture in comparison with a solid/liquid mixture at the same solid concentration; reduction of the solid/liquid interactions; self-pulverizability. thus promoting reduction of the flame length utilization of conventional burners; reduction of the flue gas flow rate; complete thermal destruction of graphite. (author)

The reaction of unirradiated and irradiated nuclear graphites with water vapor in helium

International Nuclear Information System (INIS)

Imai, Hisashi; Nomura, Shinzo; Kurosawa, Takeshi; Fujii, Kimio; Sasaki, Yasuichi

1980-10-01

Nuclear graphites more than 10 brands were oxidized with water vapor in helium and then some selected graphites were irradiated with fast neutron in the Japan Materials Testing Reactor to clarify the effect of radiation damage of graphite on their reaction behaviors. The reaction was carried out under a well defined condition in the temperature range 800 -- 1000 0 C at concentrations of water vapor 0.38 -- 1.30 volume percent in helium flow of total pressure of 1 atm. The chemical reactivity of graphite irradiated at 1000 +- 50 0 C increased linearly with neutron fluence until irradiation of 3.2 x 10 21 n/cm 2 . The activation energy for the reaction was found to decrease with neutron fluence for almost all the graphites, except for a few ones. The order of reaction increased from 0.5 for the unirradiated graphite to 1.0 for the graphite irradiated up to 6.0 x 10 20 n/cm 2 . Experiment was also performed to study a superposed effect between the influence of radiation damage of graphite and the catalytic action of barium on the reaction rate, as well as the effect of catalyser of barium. It was shown that these effects were not superposed upon each other, although barium had a strong catalytic action on the reaction. (author)

Changes in the physical and mechanical properties of graphite on irradiation in ditolylmethane

International Nuclear Information System (INIS)

Gavrilin, A.I.; Lebedev, I.G.; Sudakova, N.V.; Rizvanov, V.K.

1987-01-01

Results are presented from the irradiation and mechanical and structural testing of four grades of graphite - GMZ, VPG, MPG-6, and PG-50 - for use as moderator materials in organic cooled and graphite moderated reactors. Irradiation was carried out in the ARBUS-AST-1 reactor. Photomicrography was used to determine pore structure and ultimate strength in bending and compression was determined mechanically. Irradiation was found to increase the strength of GMZ, PMG-6, and PG-50 considerably, due to the healing of microdefects as a result of the pores filling with radiolysis products from the coolant, ditolylmethane. Conversely, VPG graphite, which has closed porosity, lost strength on irradiation

Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Krishna, R. [Dalton Cumbrian Facility, Dalton Nuclear Institute, The University of Manchester, Westlakes Science & Technology Park, Moor Row, Whitehaven, Cumbria, CA24 3HA (United Kingdom); Jones, A.N., E-mail: Abbie.Jones@manchester.ac.uk [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom); McDermott, L.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom)

2015-12-15

Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite

Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

International Nuclear Information System (INIS)

Krishna, R.; Jones, A.N.; McDermott, L.; Marsden, B.J.

2015-01-01

Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite exhibits

Irradiation growth in zirconium alloys: a review

International Nuclear Information System (INIS)

Fidleris, V.

1980-09-01

The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

Carbowaste: treatment and disposal of irradiated graphite and other carbonaceous waste

International Nuclear Information System (INIS)

Von Lensa, W.; Rizzato, C.; Baginski, K.; Banford, A.W.; Bradbury, D.; Goodwin, J.; Grambow, B.; Grave, M.J.; Jones, A.N.; Laurent, G.; Pina, G.; Vulpius, D.

2014-01-01

The European Project on 'Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)' addressed the retrieval, characterization, treatment, reuse and disposal of irradiated graphite with the following main results: - I-graphite waste features significantly depend on the specific manufacture process, on the operational conditions in the nuclear reactor (neutron dose, atmosphere, temperature etc.) and on radiolytic oxidation leading to partial releases of activation products and precursors during operation. - The neutron activation process generates significant recoil energies breaking pre-existing chemical bonds resulting in dislocations of activation products and new chemical compounds. - Most activation products exist in different chemical forms and at different locations. - I-graphite can be partly purified by thermal and chemical treatment processes leaving more leach-resistant waste products. - Leach tests and preliminary performance analyses show that i-graphite can be safely disposed of in a wide range of disposal systems, after appropriate treatment and/or conditioning. (authors)

Temperature dependence of the thermal expansion of neutron-irradiated pyrolytic carbon and graphite

International Nuclear Information System (INIS)

Matsuo, Hideto

1988-01-01

The effects of neutron irradiation and annealing on the temperature dependence of the linear thermal expansion of pyrolytic carbon and graphite were investigated after irradiation at 930-1280 0 C to a maximum neutron fluence of 2.84 x 10 25 m -2 (E > 29 fJ). After irradiation, little change in the thermal expansion of pyrolytic graphite was observed. However, as-deposited pyrolytic carbon showed an increase in thermal expansion in the perpendicular direction, a decrease in the direction parallel to the deposition plane, and also an increase in the anisotropy of the thermal expansion. Annealing at 2000 0 C did not cause any effective changes for irradiated specimens of either as-deposited pyrolytic carbon or pyrolytic graphite. (author)

Ion irradiation of {sup 37}Cl implanted nuclear graphite: Effect of the energy deposition on the chlorine behavior and consequences for the mobility of {sup 36}Cl in irradiated graphite

Energy Technology Data Exchange (ETDEWEB)

Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Moncoffre, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Sainsot, P. [Université de Lyon, Université Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259 (France); Rouzaud, J.-N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure (ENS), Paris, UMR CNRS-ENS 8538 (France)

2015-09-15

Graphite is used in many types of nuclear reactors due to its ability to slow down fast neutrons without capturing them. Whatever the reactor design, the irradiated graphite waste management has to be faced sooner or later regarding the production of long lived or dose determining radioactive species such as {sup 14}C, {sup 3}H or {sup 36}Cl. The first carbon dioxide cooled, graphite moderated nuclear reactors resulted in a huge quantity of irradiated graphite waste for which the management needs a previous assessment of the radioactive inventory and the radionuclide’s location and speciation. As the detection limits of usual spectroscopic methods are generally not adequate to detect the low concentration levels (<1 ppm) of the radionuclides, we used an indirect approach based on the implantation of {sup 37}Cl, to simulate the presence of {sup 36}Cl. Our previous studies show that temperature is one of the main factors to be considered regarding the structural evolution of nuclear graphite and chlorine mobility during reactor operation. However, thermal release of chlorine cannot be solely responsible for the depletion of the {sup 36}Cl inventory. We propose in this paper to study the impact of irradiation and its synergetic effects with temperature on chlorine release. Indeed, the collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic collisions. However, a small part of the recoil carbon atom energy is also transferred to the lattice through electronic excitation. This paper aims at elucidating the effects of the different irradiation regimes (ballistic and electronic) using ion irradiation, on the mobility of implanted {sup 37}Cl, taking into account the initial disorder level of the nuclear graphite.

Verification of thermal-irradiation stress analytical code VIENUS of graphite block

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

1992-02-01

The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

Rate-dependent mode I interlaminar crack growth mechanisms in graphite/epoxy and graphite/PEEK

Science.gov (United States)

Gillespie, J. W., Jr.; Carlsson, L. A.; Smiley, A. J.

1987-01-01

In this paper the mode I fracture behavior of graphite/epoxy and graphite/PEEK composites is examined over four decades of crosshead rates (0.25-250 mm/min). Straight-sided double-cantilever-beam specimens consisting of unidirectional laminates were tested at room temperature. For graphite/epoxy the load-deflection response was linear to fracture, and stable slow crack growth initiating at the highest load level was observed for all rates tested. In contrast, mode I crack growth in the graphite/PEEK material was often unstable and showed stick-slip behavior. Subcritical crack growth occurring prior to the onset of fracture was observed at intermediate displacement rates. A mechanism for the fracture behavior of the graphite/PEEK material (based on viscoelastic, plastic, and microcrack coalescence in the process zone) is proposed and related to the observed rate-dependent phenomena.

Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

1976-01-01

A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

International Nuclear Information System (INIS)

2016-05-01

Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

Formation of dislocation dipoles in irradiated graphite

International Nuclear Information System (INIS)

Niwase, Keisuke

2005-01-01

Recently, we have proposed a dislocation dipole accumulation model to explain the irradiation-induced amorphization of graphite. However, the structure of dislocation dipole in the hexagonal networks is still an open question at the atomic-level. In this paper, we propose a possible formation process of the dislocation dipole

Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

Directory of Open Access Journals (Sweden)

Xuegang Liu

2017-01-01

Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

Thermal conductivity degradation of graphites due to neutron irradiation at low temperature

International Nuclear Information System (INIS)

Snead, L.L.; Burchell, T.D.

1995-01-01

Several graphites and carbon/carbon composites (C/C's) have been irradiated with fission neutrons near 150 C and at fluences up to a displacement level of 0.24 dpa. The unirradiated room temperature thermal conductivity of these materials varied from 114 W/m K for H-451 isotropic graphite, to 670 W/m K for a unidirectional FMI-1D C/C composite. At the irradiation temperature a saturation reduction in thermal conductivity was seen to occur at displacement levels of approximately 0.1 dpa. All materials were seen to degrade to approximately 10 to 14% of their original thermal conductivity after irradiation. The significant recovery of thermal conductivity due to post-irradiation isochronal anneals is also presented. (orig.)

Polymer surfaces graphitization by low-energy He{sup +} ions irradiation

Energy Technology Data Exchange (ETDEWEB)

Geworski, A.; Lazareva, I.; Gieb, K.; Koval, Y.; Müller, P. [Department of Physics, Universität Erlangen-Nürnberg, Erwin-Rommel-Str. 1, 91058 Erlangen (Germany)

2014-08-14

The electrical and optical properties of surfaces of polyimide and AZ5214e graphitized by low-energy (1 keV) He{sup +} irradiation at different polymer temperatures were investigated. The conductivity of the graphitized layers can be controlled with the irradiation temperature within a broad range and can reach values up to ∼1000 S/cm. We show that the electrical transport in low-conducting samples is governed by thermally activated hopping, while the samples with a high conductivity show a typical semimetallic behavior. The transition from thermally activated to semimetallic conductance governed by the irradiation temperature could also be observed in optical measurements. The semimetallic samples show an unusually high for graphitic materials carrier concentration, which results in a high extinction coefficient in the visible light range. By analyzing the temperature dependence of the conductance of the semimetallic samples, we conclude that the scattering of charge carriers is dominated by Coulomb interactions and can be described by a weak localization model. The transition from a three to two dimensional transport mechanism at low temperatures consistently explains the change in the temperature dependence of the conductance by cooling, observed in experiments.

Special graphites; Graphites speciaux

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

Irradiation-induced defects in graphite and glassy carbon studied by positron annihilation

International Nuclear Information System (INIS)

Hasegawa, M.; Kajino, M.; Kuwahara, H.; Yamaguchi, S.; Kuramoto, E.; Takenaka, M.

1992-01-01

ACAR and positron lifetime measurements have been made on, HOPG, isotropic fine-grained graphites, glassy carbons and C 60 /C 70 . HOPG showed a marked bimodal ACAR distribution along the c-axis. By irradiation of 1.0 X 10 19 fast neutrons/cm 2 remarkable narrowing in the ACAR curves and disappearance of the bimodal distribution were observed. Lifetime in HOPG increased from 225 psec to 289 psec (positron-lifetime in vacancies and their small clusters) by the irradiation. The irradiation on isotropic graphites and glassy carbons, however, gave slight narrowing in ACAR curves and decrease in lifetimes (360 psec → 300psec). This suggests irradiation-induced vacancy trapping in crystallites. In C 60 /C 70 powder two lifetime components were detected: τ 1 =177psec, τ 2 =403psec (I 2 =58%). The former is less than the bulk lifetime of HOPG, while the latter being very close to lifetimes in the isotropic graphites and glassy carbons. This and recent 2D-ACAR study of HOPG surface [15] strongly suggest free and defect surface states around ''soccer ball'' cages

Differences in the irradiation effects of IG-110 and IG-430 nuclear graphites : effects of coke difference

International Nuclear Information System (INIS)

Chi, Se Hwan; Kim, Gen Chan; Kim, Eung Seon; Hong, Jin Ki; Chang, Jong Hwa

2005-01-01

In the high temperature gas cooled reactors (HTGRs), graphite acts as a moderator and reflector as well as a major structural component that may provide channels for the fuel and coolant gas, channels for control and shut down, and thermal and neutron shielding. During a reactor operation, many of the physical, chemical and mechanical properties of these graphite components are significantly modified as a function of the temperature, environment, and an irradiation. On the other hand, currently, all the nuclear graphites are being manufactured from two types of cokes, i.e., petroleum and coal-tar pitch coke, and it has been understood that the type of coke plays the most critical role determining the properties of a specific graphite grade. To investigate the effects of coke types on the irradiation response of a graphite, two graphites of different cokes were irradiated by 3 MeV C+ ions and the differences in the response of ion-irradiation were investigated

The behavior of interstitials in irradiated graphite

International Nuclear Information System (INIS)

Pedraza, D.F.

1991-01-01

A computer model is developed to simulate the behavior of self-interstitials with particular attention to clustering. Owing to the layer structure of graphite, atomistic simulations can be performed using a large parallelepipedic supercell containing a few layers. In particular, interstitial clustering is studied here using a supercell that contains two basal planes only. Frenkel pairs are randomly produced. Interstitials are placed at sites between the crystal planes while vacancies are distributed in the two crystal planes. The size of the computational cell is 20000 atoms and periodic boundary conditions are used in two dimensions. Vacancies are assumed immobile whereas interstitials are given a certain mobility. Two point defect sinks are considered, direct recombination of Frenkel pairs and interstitial clusters. The clusters are assumed to be mobile up to a certain size where they are presumed to become loop nuclei. Clusters can shrink by emission of singly bonded interstitials or by recombination of a peripheral interstitial with a neighboring vacancy. The conditions under which interstitial clustering occurs are reported. It is shown that when clustering occurs the cluster size population gradually shifts towards the largest size cluster. The implications of the present results for irradiation growth and irradiation-induced amorphization are discussed

Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

Energy Technology Data Exchange (ETDEWEB)

Everett, M. R.; Graham, L. W.; Ridealgh, F.

1971-11-15

Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.

Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

Science.gov (United States)

Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

1995-08-01

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

Change in physical properties of high density isotropic graphites irradiated in the ''JOYO'' fast reactor

International Nuclear Information System (INIS)

Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

1995-01-01

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor ''JOYO'' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597 C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (orig.)

Irradiation damage in graphite. The works of Professor B.T. Kelly

International Nuclear Information System (INIS)

Marsden, B.J.

1996-01-01

The irradiation damage produced in graphite by energetic neutrons (>100eV) has been extensively studied because of the use of graphite as a moderator in thermal nuclear reactors. In recent times, graphite has been adopted as the protective tiling of the inner wall of experimental fusion systems and property changes due to fusion neutrons have become important. The late Professor B.T. Kelly reviewed the work carried out on the irradiation behaviour of graphite since the 1940s. This work is particularly timely as the scale of research into the effects of fission neutrons has been greatly reduced and many of the active researchers have retired. In recent years, new programmes of work are being formulated for the use of graphite in both the field of high temperature reactor systems and fusion systems. It is therefore important that the knowledge gained by Professor Kelly and other workers is not lost but passed on to future generations of nuclear scientists and engineers. This paper reviews Professor Kelly's last work, it also draws on the experience gained during many long discussions with Brian during the years he worked closely with the present graphite team at AEA Technology. It is hoped to publish his work in full in the near future. (author). 13 refs, 14 figs, 3 tabs

Studies on the graphite rupture under irradiation induced strains

International Nuclear Information System (INIS)

Jouquet, G.; Berthion, Y.; L'Homme, A.

1980-01-01

Following the RMG experiments (failure of graphite by mechanical effect, i.e. under very high temperature gradient) an experimental program called RWG (Failure of Graphite by WIGNER effect) was initiated in 75 at C.E.A. 3 experiments have been already performed in the OSIRIS reactor at Saclay: RWG 01, 02 and 03. A 4th one, RWG04, is scheduled for the end of 79, may be in collaboration with GERMANY. The aim of the RWG experiments is to induce internal stresses in graphite blocks by irradiation at high temperature which would lead or not to their failure so one could bracket, as tightly as possible, the critical value for failure onset in given experimental conditions

Characterization of {sup 14}C in neutron irradiated NBG-25 nuclear graphite

Energy Technology Data Exchange (ETDEWEB)

LaBrier, Daniel, E-mail: labrdani@isu.edu; Dunzik-Gougar, Mary Lou

2014-05-01

Recent studies suggest that the highest concentration of {sup 14}C contamination present in reactor-irradiated graphite exists on the surfaces and within near-surface layers. Surface-sensitive analysis techniques (XPS, ToF-SIMS, SEM/EDS and Raman) were employed to determine the chemical nature of {sup 14}C on irradiated NBG-25 (nuclear grade) graphite surfaces. Several {sup 14}C precursor species are identified on the surfaces of irradiated NBG-25; the quantities of these species decrease at sub-surface depths, which further suggests that {sup 14}C formation is predominantly a surface-concentrated phenomenon. The elevated presence of several surface oxide complexes on irradiated NBG-25 surfaces are attributed directly to neutron irradiation. Larger numbers of oxide bonds were found on irradiated NBG-25 surfaces (when compared to unirradiated samples) in the form of interlattice (e.g. ether) and dangling (e.g. carboxylate and ketone) bonds; the quantities of these bond types also decrease with increasing sub-surface depths.

Swift heavy ions induced irradiation effects in monolayer graphene and highly oriented pyrolytic graphite

International Nuclear Information System (INIS)

Zeng, J.; Yao, H.J.; Zhang, S.X.; Zhai, P.F.; Duan, J.L.; Sun, Y.M.; Li, G.P.; Liu, J.

2014-01-01

Monolayer graphene and highly oriented pyrolytic graphite (HOPG) were irradiated by swift heavy ions ( 209 Bi and 112 Sn) with the fluence between 10 11 and 10 14 ions/cm 2 . Both pristine and irradiated samples were investigated by Raman spectroscopy. It was found that D and D′ peaks appear after irradiation, which indicated the ion irradiation introduced damage both in the graphene and graphite lattice. Due to the special single atomic layer structure of graphene, the irradiation fluence threshold Φ th of the D band of graphene is significantly lower ( 11 ions/cm 2 ) than that (2.5 × 10 12 ions/cm 2 ) of HOPG. The larger defect density in graphene than in HOPG indicates that the monolayer graphene is much easier to be damaged than bulk graphite by swift heavy ions. Moreover, different defect types in graphene and HOPG were detected by the different values of I D /I D′ . For the irradiation with the same electronic energy loss, the velocity effect was found in HOPG. However, in this experiment, the velocity effect was not observed in graphene samples irradiated by swift heavy ions

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

Science.gov (United States)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2018-04-01

Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

Analytical and numerical study of graphite IG110 parts in advanced reactor under high temperature and irradiation

Energy Technology Data Exchange (ETDEWEB)

Gao, Jinling, E-mail: Jinling_Gao@yeah.net; Yao, Wenjuan, E-mail: wj_yao@yeah.net; Ma, Yudong

2016-08-15

Graphical abstract: An analytical model and a numerical procedure are developed to study the mechanical response of IG-110 graphite bricks in HTGR subjected to high temperature and irradiation. The calculation results show great accordance with each other. Rational suggestions on the calculation and design of the IG-110 graphite structure are proposed based on the sensitivity analyses including temperature, irradiation dimensional change, creep and Poisson’s ratio. - Highlights: • Analytical solution of stress and displacement of IG-110 graphite components in HTGR. • Finite element procedure developed for stress analysis of HTGR graphite component. • Parameters analysis of mechanical response of graphite components during the whole life of the reflector. - Abstract: Structural design of nuclear power plant project is an important sub-discipline of civil engineering. Especially after appearance of the fourth generation advanced high temperature gas cooled reactor, structural mechanics in reactor technology becomes a popular subject in structural engineering. As basic ingredients of reflector in reactor, graphite bricks are subjected to high temperature and irradiation and the stress field of graphite structures determines integrity of reflector and makes a great difference to safety of whole structure. In this paper, based on assumptions of elasticity, side reflector is regarded approximately as a straight cylinder structure and primary creep strain is ignored. An analytical study on stress of IG110 graphite parts is present. Meanwhile, a finite element procedure for calculating stresses in the IG110 graphite structure exposed in the high temperature and irradiation is developed. Subsequently, numerical solution of stress in IG110 graphite structure is obtained. Analytical solution agrees well with numerical solution, which indicates that analytical derivation is accurate. Finally, influence of temperature, irradiation dimensional change, creep and Poisson�

Inert annealing of irradiated graphite by inductive heating

International Nuclear Information System (INIS)

Botzem, W.; Woerner, J.

2001-01-01

Fission neutrons change physical properties of graphite being used in nuclear reactors as moderator and as structural material. The understanding of these effects on an atomic model is expressed by dislocations of carbon atoms within the graphite and the thereby stored energy is known as Wigner Energy. The dismantling of the Pile 1 core may necessitate the thermal treatment of the irradiated but otherwise undamaged graphite. This heat treatment - usually called annealing - initiates the release of stored Wigner Energy in a controlled manner. This energy could otherwise give rise to an increase in temperature under certain conditions during transport or preparation for final storage. In order to prevent such an effect it is intended to anneal the major part of Pile 1 graphite before it is packed into boxes suitable for final disposal. Different heating techniques have been assessed. Inductive heating in an inert atmosphere was selected for installation in the Pile 1 Waste Processing Facility built for the treatment and packaging of the dismantled Pile 1 waste. The graphite blocks will be heated up to 250 deg. C in the annealing ovens, which results in the release of significant amount of the stored energy. External heat sources in a final repository will never heat up the storage boxes to such a temperature. (author)

Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

International Nuclear Information System (INIS)

2016-05-01

Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

Nucleation and growth characteristics of graphite spheroids in bainite during graphitization annealing of a medium carbon steel

International Nuclear Information System (INIS)

Gao, J.X.; Wei, B.Q.; Li, D.D.; He, K.

2016-01-01

The evolution of microstructure in bainite during graphitization annealing at 680 °C of Jominy-quenched bars of an Al-Si bearing medium carbon (0.4C wt%) steel has been studied and compared with that in martensite by using light, scanning and transmission electron microscopy. The results show that the graphitization process in bainite is different from that in martensite in many aspects such as the initial carbon state, the behavior of cementite, the nucleation-growth feature and kinetics of formation of graphite spheroids during graphitization annealing, and the shape, size and distribution of these graphite spheroids. The fact that the graphitization in bainite can produce more homogeneous graphite spheroids with more spherical shape and finer size in a shorter annealing time without the help of preexisting coring particles implies that bainite should be a better starting structure than martensite for making graphitic steel. - Highlights: • This article presents a microstructural characterization of formation of graphite spheroids in bainite. • Nucleation and growth characteristics of graphite spheroids formed in bainite and martensite are compared. • Bainite should be a better starting structure for making graphitic steel as results show.

Study by internal friction of curing low temperature irradiation defects in graphite

International Nuclear Information System (INIS)

Rouby, Dominique.

1974-01-01

Micromechanical properties and anelastic effects of neutrons irradiated graphites at 300 and 77 0 K are investigated by internal friction analysis and elasticity modulus variations. Defects created by irradiation are studied and evolution versus dose and annealing is followed [fr

IAEA international database on irradiated nuclear graphite properties

International Nuclear Information System (INIS)

Burchell, T.D.; Clark, R.E.H.; Stephens, J.A.; Eto, M.; Haag, G.; Hacker, P.; Neighbour, G.B.; Janev, R.K.; Wickham, A.J.

2000-02-01

This report describes an IAEA database containing data on the properties of irradiated nuclear graphites. Development and implementation of the graphite database followed initial discussions at an IAEA Specialists' Meeting held in September 1995. The design of the database is based upon developments at the University of Bath (United Kingdom), work which the UK Health and Safety Executive initially supported. The database content and data management policies were determined during two IAEA Consultants' Meetings of nuclear reactor graphite specialists held in 1998 and 1999. The graphite data are relevant to the construction and safety case developments required for new and existing HTR nuclear power plants, and to the development of safety cases for continued operation of existing plants. The database design provides a flexible structure for data archiving and retrieval and employs Microsoft Access 97. An instruction manual is provided within this document for new users, including installation instructions for the database on personal computers running Windows 95/NT 4.0 or higher versions. The data management policies and associated responsibilities are contained in the database Working Arrangement which is included as an Appendix to this report. (author)

Quality assurance for the IAEA International Database on Irradiated Nuclear Graphite Properties

International Nuclear Information System (INIS)

Wickham, A.J.; Humbert, D.

2006-06-01

Consideration has been given to the process of Quality Assurance applied to data entered into current versions of the IAEA International Database on Irradiated Nuclear Graphite Properties. Originally conceived simply as a means of collecting and preserving data on irradiation experiments and reactor operation, the data are increasingly being utilised for the preparation of safety arguments and in the design of new graphites for forthcoming generations of graphite-moderated plant. Under these circumstances, regulatory agencies require assurances that the data are of appropriate accuracy and correctly transcribed, that obvious errors in the original documentation are either highlighted or corrected, etc., before they are prepared to accept analyses built upon these data. The processes employed in the data transcription are described in this document, and proposals are made for the categorisation of data and for error reporting by Database users. (author)

Modification of graphite structure by irradiation, revealed by thermal oxidation. Examination by electronic microscopy

International Nuclear Information System (INIS)

Rouaud, Michel

1969-01-01

Based on the analysis of images obtained by electronic microscopy, this document reports the comparative study of the action of neutrons on three different graphites: a natural one (Ticonderoga) and two pyrolytic ones (Carbone-Lorraine and Raytheon). The approach is based on the modification of features of thermal oxidation of graphites by dry air after irradiation. Different corrosion features are identified. The author states that there seems to be a relationship between the number and shape of these features, and defects existing on the irradiated graphite before oxidation. For low doses, the feature aspect varies with depth at which oxidation occurs. For higher doses, the aspect remains the same [fr

Exercise of Intercomparison on characterization of graphite irradiated: CW-RRT

International Nuclear Information System (INIS)

Pina, G.; Rodriguez, M.; Lara, E.; Magro, E.

2014-01-01

Knowledge of inventory of irradiated graphite (i-graphite) of nuclear reactors is a critical parameter for the proper management of waste. For this reason it is of paramount importance have access to reliable analytical methodologies to provide accurate results through radiochemical procedures specifics. La reliability of the result depends on both its precision and the accuracy obtained. this requires specific tasks compared to other methods, such as the extent of an array reference or conducting intercomparison exercises between laboratories. (Author)

Development of fracture toughness test method for nuclear grade graphite

Energy Technology Data Exchange (ETDEWEB)

Chang, C. H.; Lee, J. S.; Cho, H. C.; Kim, D. J.; Lee, D. J. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2006-02-15

Because of its high strength and stability at very high temperature, as well as very low thermal neutron absorption cross-section, graphite has been widely used as a structural material in Gas Cooled Reactors (GCR). Recently, many countries are developing the Very High Temperature gas cooled Reactor (VHTR) because of the potentials of hydrogen production, as well as its safety and viable economics. In VHTR, helium gas serves as the primary coolant. Graphite will be used as a reflector, moderator and core structural materials. The life time of graphite is determined from dimensional changes due to neutron irradiation, which closely relates to the changes of crystal structure. The changes of both lattice parameter and crystallite size can be easily measured by X-ray diffraction method. However, due to high cost and long time of neutron irradiation test, ion irradiation test is being performed instead in KAERI. Therefore, it is essential to develop the technique for measurement of ion irradiation damage of nuclear graphite. Fracture toughness of nuclear grade graphite is one of the key properties in the design and development of VHTR. It is important not only to evaluate the various properties of candidate graphite but also to assess the integrity of nuclear grade graphite during operation. Although fracture toughness tests on graphite have been performed in many laboratories, there have been wide variations in values of the calculated fracture toughness, due to the differences in the geometry of specimens and test conditions. Hence, standard test method for nuclear graphite is required to obtain the reliable fracture toughness values. Crack growth behavior of nuclear grade graphite shows rising R-curve which means the increase in crack growth resistance as the crack length increases. Crack bridging and microcracking have been proposed to be the dominant mechanisms of rising R-curve behavior. In this paper, the technique to measure the changes of crystallite size and

High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

Energy Technology Data Exchange (ETDEWEB)

Manzel, R.; Everett, M. R.; Graham, L. W.

1971-05-15

The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 10²¹ n cm^-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

Study on "1"4C content in post-irradiation graphite spheres of HTR-10

International Nuclear Information System (INIS)

Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

2014-01-01

Since the production mechanism of the "1"4C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of "1"4C in fuel elements and graphite spheres of HTR-10, the production mechanism of the "1"4C in graphite spheres was studied. The production sources of the "1"4C in graphite spheres and fuel elements were summarized, the amount of "1"4C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the "1"4C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

International Nuclear Information System (INIS)

Arsene, Carmen

2009-01-01

The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)

Misorientations in spheroidal graphite: some new insights about spheroidal graphite growth in cast irons

International Nuclear Information System (INIS)

Lacaze, J; Theuwissen, K; Laffont, L; Véron, M

2016-01-01

Local diffraction patterning, orientation mapping and high resolution transmission electron microscopy imaging have been used to characterize misorientations in graphite spheroids of cast irons. Emphasis is put here on bulk graphite, away from the nucleus as well as from the outer surface of the spheroids in order to get information on their growth during solidification. The results show that spheroidal graphite consists in conical sectors made of elementary blocks piled up on each other. These blocks are elongated along the prismatic a direction of graphite with the c axes roughly parallel to the radius of the spheroids. This implies that the orientation of the blocks rotates around the spheroid centre giving low angle tilting misorientations along tangential direction within each sector. Misorientations between neighbouring sectors are of higher values and their interfaces show rippled layers which are characteristic of defects in graphene. Along a radius of the spheroid, clockwise and anticlockwise twisting between blocks is observed. These observations help challenging some of the models proposed to explain spheroidal growth in cast ions. (paper)

The irradiation creep in reactor graphites for HTR applications

International Nuclear Information System (INIS)

Veringa, H.J.; Blackstone, R.

1976-01-01

A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400 0 C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400 0 C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density. (author)

Graphites and composites irradiations for gas cooled reactor core structures

International Nuclear Information System (INIS)

Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

2008-01-01

Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

An Electron Microscopy Study of Graphite Growth in Nodular Cast Irons

Science.gov (United States)

Laffont, L.; Jday, R.; Lacaze, J.

2018-04-01

Growth of graphite during solidification and high-temperature solid-state transformation has been investigated in samples cut out from a thin-wall casting which solidified partly in the stable (iron-graphite) and partly in the metastable (iron-cementite) systems. Transmission electron microscopy has been used to characterize graphite nodules in as-cast state and in samples having been fully graphitized at various temperatures in the austenite field. Nodules in the as-cast material show a twofold structure characterized by an inner zone where graphite is disoriented and an outer zone where it is well crystallized. In heat-treated samples, graphite nodules consist of well-crystallized sectors radiating from the nucleus. These observations suggest that the disoriented zone appears because of mechanical deformation when the liquid contracts during its solidification in the metastable system. During heat-treatment, the graphite in this zone recrystallizes. In turn, it can be concluded that nodular graphite growth mechanism is the same during solidification and solid-state transformation.

Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

Energy Technology Data Exchange (ETDEWEB)

Pinhero, Patrick [Univ. of Missouri, Columbia, MO (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-10

The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase, i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be

An evaluation on fatigue crack growth in a fine-grained isotropic graphite

International Nuclear Information System (INIS)

Wang Hongtao; Sun Libin; Li Chenfeng; Shi Li; Wang Haitao

2012-01-01

Highlights: ► The propagation of micro- and macro-fatigue cracks in IG-11 graphite was studied. ► The curves of the fatigue crack growth rate versus the SIF range show three stages. ► The fatigue microcrack propagation is very sensitive to graphite's microstructures. ► Graphite's microstructures have no significant impact on fatigue macrocrack growth. ► The fatigue fracture surface indicates the fracture mechanism of the IG-11 graphite. - Abstract: The aim of this paper is to investigate the mechanism of fatigue crack propagation in IG-11 graphite, and determine the crack growth rate in relation to the stress level. Experimental studies were performed at both micro and macro scales. For fatigue microcrack propagation, single-edge-notch specimens were chosen for testing and the fatigue crack growth was measured in situ with a scanning electron microscope. For fatigue macrocrack propagation, CT specimens were used and the fatigue crack growth was measured with a high-accuracy optic microscope. Combining the two groups of experimental results, the following conclusions are derived: (1) The heterogeneous microstructures of the graphite material have significant impact on the fatigue microcrack growth, while their influence on fatigue macrocrack growth is very limited. (2) The relationship between the fatigue crack growth rate and the crack-tip stress intensity factor range can be expressed in the form of Paris formulae, which contains three stages: an initial rising part with a small slope, an abrupt rise with a very large acceleration, and a short final part with a small slope. (3) The fatigue fracture surface of the graphite material contains considerable sliding of leaf-shape graphite flakes combined with small cotton-shape plastic deformations. These sliding traces are approximately aligned at 45°, showing the main cause of the fatigue fracture is the shear stress. There are also a large amount of secondary cracks inside unit cells and on cell walls

A reverse method for the determination of the radiological inventory of irradiated graphite at reactor scale

Energy Technology Data Exchange (ETDEWEB)

Nicaise, Gregory [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-roses (France); Poncet, Bernard [EDF-DP2D, Lyon (France)

2016-11-15

Irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF). Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Because of graphite high purity, radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. Therefore the assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A reverse method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible.

Graphite selection for the PBMR reflector

International Nuclear Information System (INIS)

Marsden, B.J.; Preston, S.D.

2000-01-01

A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

International Nuclear Information System (INIS)

Anfimov, S.S.

2000-01-01

As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

Fabrication of SnO2-Reduced Graphite Oxide Monolayer-Ordered Porous Film Gas Sensor with Tunable Sensitivity through Ultra-Violet Light Irradiation

Science.gov (United States)

Xu, Shipu; Sun, Fengqiang; Yang, Shumin; Pan, Zizhao; Long, Jinfeng; Gu, Fenglong

2015-01-01

A new graphene-based composite structure, monolayer-ordered macroporous film composed of a layer of orderly arranged macropores, was reported. As an example, SnO2-reduced graphite oxide monolayer-ordered macroporous film was fabricated on a ceramic tube substrate under the irradiation of ultra-violet light (UV), by taking the latex microsphere two-dimensional colloid crystal as a template. Graphite oxide sheets dispersed in SnSO4 aqueous solution exhibited excellent affinity with template microspheres and were in situ incorporated into the pore walls during UV-induced growth of SnO2. The growing and the as-formed SnO2, just like other photocatalytic semiconductor, could be excited to produce electrons and holes under UV irradiation. Electrons reduced GO and holes adsorbed corresponding negative ions, which changed the properties of the composite film. This film was directly used as gas-sensor and was able to display high sensitivity in detecting ethanol gas. More interestingly, on the basis of SnO2-induced photochemical behaviours, this sensor demonstrated tunable sensitivity when UV irradiation time was controlled during the fabrication process and post in water, respectively. This study provides efficient ways of conducting the in situ fabrication of a semiconductor-reduced graphite oxide film device with uniform surface structure and controllable properties. PMID:25758292

Leaching of 14C and 36Cl from irradiated French graphite

International Nuclear Information System (INIS)

Gray, W.J.; Morgan, W.C.

1989-08-01

The leach rates of 14 C and 36 Cl were measured on solid cylindrical samples prepared from irradiated graphite blocks supplied by the French Commissariat a l'Energie Atomique (CEA). Static leach tests were conducted in deionized water at 20 degree C for 13 weeks. The graphite samples were completely submerged in the water, and the entire volume of water was changed and analyzed at weekly intervals for the first three weeks and biweekly thereafter. Large differences in the leach rates of both 14 C and 36 Cl were observed between samples machined from the different blocks. In general, the leach rates were much higher than those measured in an earlier study with graphite obtained from a block removed from one of the Hanford reactors. The data from this study are compared with those from the previous study using the Hanford-reactor graphite. Implications of the data from both studies regarding possible rate-limiting mechanisms are discussed. 4 refs., 8 figs., 3 tabs

An evaluation on fatigue crack growth in a fine-grained isotropic graphite

Energy Technology Data Exchange (ETDEWEB)

Wang Hongtao; Sun Libin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Li Chenfeng [College of Engineering, Swansea University, Singleton Park, Swansea SA2 8PP (United Kingdom); Shi Li [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Wang Haitao, E-mail: wanght@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

2012-09-15

Highlights: Black-Right-Pointing-Pointer The propagation of micro- and macro-fatigue cracks in IG-11 graphite was studied. Black-Right-Pointing-Pointer The curves of the fatigue crack growth rate versus the SIF range show three stages. Black-Right-Pointing-Pointer The fatigue microcrack propagation is very sensitive to graphite's microstructures. Black-Right-Pointing-Pointer Graphite's microstructures have no significant impact on fatigue macrocrack growth. Black-Right-Pointing-Pointer The fatigue fracture surface indicates the fracture mechanism of the IG-11 graphite. - Abstract: The aim of this paper is to investigate the mechanism of fatigue crack propagation in IG-11 graphite, and determine the crack growth rate in relation to the stress level. Experimental studies were performed at both micro and macro scales. For fatigue microcrack propagation, single-edge-notch specimens were chosen for testing and the fatigue crack growth was measured in situ with a scanning electron microscope. For fatigue macrocrack propagation, CT specimens were used and the fatigue crack growth was measured with a high-accuracy optic microscope. Combining the two groups of experimental results, the following conclusions are derived: (1) The heterogeneous microstructures of the graphite material have significant impact on the fatigue microcrack growth, while their influence on fatigue macrocrack growth is very limited. (2) The relationship between the fatigue crack growth rate and the crack-tip stress intensity factor range can be expressed in the form of Paris formulae, which contains three stages: an initial rising part with a small slope, an abrupt rise with a very large acceleration, and a short final part with a small slope. (3) The fatigue fracture surface of the graphite material contains considerable sliding of leaf-shape graphite flakes combined with small cotton-shape plastic deformations. These sliding traces are approximately aligned at 45 Degree-Sign , showing the

Nuclear graphite based on coal tar pitch; behavior under neutron irradiation between 400 and 14000C

International Nuclear Information System (INIS)

Mottet, P.; Fillatre, A.; Schill, R.; Micaud, G.

1977-01-01

Two nuclear grades of coal tar pitch coke graphites have been developed and tested under neutron irradiation. The neutron irradiation induced dimensional changes between 400 and 1400 0 C, at fluences up to 1,2.10 22 n.cm -2 PHI.FG show a behavior comparable to anisotropic petroleum coke graphites. Less than 10% variation in thermal expansion, maximum decrease by a factor four in thermal conductivity, and large increase of the Young modulus have been observed

Carbon Nanotubes Growth by CVD on Graphite Fibers

Science.gov (United States)

Zhu, Shen; Su, Ching-Hua; Cochrane, J. C.; Lehoczky, S. L.; Muntele, I.; Ila, D.; Curreri, Peter A. (Technical Monitor)

2002-01-01

Due to the superior electrical and mechanical properties of carbon nanotubes (CNT), synthesizing CNT on various substances for electronics devices and reinforced composites have been engaged in many efforts for applications. This presentation will illustrate CNT synthesized on graphite fibers by thermal CVD. On the fiber surface, iron nanoparticles as catalysts for CNT growth are coated. The growth temperature ranges from 600 to 1000 C and the pressure ranges from 100 Torr to one atmosphere. Methane and hydrogen gases with methane content of 10% to 100% are used for the CNT synthesis. At high growth temperatures (greater than or equal to 900 C), the rapid inter-diffusion of the transition metal iron on the graphite surface results in the rough fiber surface without any CNT grown on it. When the growth temperature is relative low (650-800 C), CNT with catalytic particles on the nanotube top ends are fabricated on the graphite surface. (Methane and hydrogen gases with methane content of 10% to 100% are used for the CNT synthesis.) (By measuring the samples) Using micro Raman spectroscopy in the breath mode region, single-walled or multi-walled CNT (MWCNT), depending on growth concentrations, are found. Morphology, length and diameter of these MWCNT are determined by scanning electron microscopy and Raman spectroscopy. The detailed results of syntheses and characterizations will be discussed in the presentation.

Irradiation damage in graphite due to fast neutrons in fission and fusion systems

International Nuclear Information System (INIS)

2000-09-01

Gas cooled reactors have been in operation for the production of electricity for over forty years, encompassing a total of 56 units operated in seven countries. The predominant experience has been with carbon dioxide cooled reactors (52 units), with the majority operated in the United Kingdom. In addition, four prototype helium cooled power plants were operated in the United States and Germany. The United Kingdom has no plans for further construction of carbon dioxide units, and the last helium cooled unit was shutdown in 1990. However, there has been an increasing interest in modular helium cooled reactors during the 1990s as a possible future nuclear option. Graphite is a primary material for the construction of gas cooled reactor cores, serving as a low absorption neutron moderator and providing a high temperature, high strength structure. Commercial gas cooled reactor cores (both carbon dioxide cooled and helium cooled) utilise large quantities of graphite. The structural behaviour of graphite (strength, dimensional stability, susceptibility to cracking, etc.) is a complex function of the source material, manufacturing process, chemical environment, and temperature and irradiation history. A large body of data on graphite structural performance has accumulated from operation of commercial gas cooled reactors, beginning in the 1950s and continuing to the present. The IAEA is supporting a project to collect graphite data and archive it in a retrievable form as an International Database on Irradiated Nuclear Graphite Properties, with limited general access and more detailed access by participating Member States. Because of the large size of the database, the complexity of the phenomena and the number of variables involved, a general understanding of graphite behaviour is essential to the understanding and use of the data

Negative pressure and spallation in graphite targets under nano- and picosecond laser irradiation

Energy Technology Data Exchange (ETDEWEB)

Belikov, R S; Khishchenko, K V [Joint Institute for High Temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Krasyuk, I K; Semenov, A Yu; Stuchebryukhov, I A [A M Prokhorov General Physics Institute, Russian Academy of Sciences, Moscow (Russian Federation); Rinecker, T; Schoenlein, A [Goethe University Frankfurt am Main (Germany); Rosmej, O N [GSI Helmholtzzentrum für Schwerionenforschung GmbH, Germany, 64291 Darmstadt, Planckstraße, 1 (Germany); Tomut, M [Technische Universität Darmstadt, Germany, 64289 Darmstadt, Karolinenplatz, 5 (Germany)

2015-05-31

We present the results of experiments on the spallation phenomena in graphite targets under shock-wave nano- and picosecond irradiation, which have been performed on Kamerton-T (GPI, Moscow, Russia) and PHELIX (GSI, Darmstadt, Germany) laser facilities. In the range of the strain rates of 10{sup 6} – 10{sup 7} s{sup -1}, the data on the dynamic mechanical strength of the material at rapure (spallation) have been for the first time obtained. With a maximal strain rate of 1.4 × 10{sup 7} s{sup -1}, the spall strength of 2.1 GPa is obtained, which constitutes 64% of the theoretical ultimate tensile strength of graphite. The effect of spallation is observed not only on the rear side of the target, but also on its irradiated (front) surface. With the use of optical and scanning electron microscopes, the morphology of the front and rear surfaces of the targets is studied. By means of Raman scattering of light, the graphite structure both on the target front surface under laser exposure and on its rear side in the spall zone is investigated. A comparison of the dynamic strength of graphite and synthetic diamond is performed. (extreme light fields and their applications)

Determining the future for irradiated graphite disposal

International Nuclear Information System (INIS)

Neighbour, G.B.; Wickham, A.J.; Hacker, P.J.

2000-01-01

In recent years, proposals have been made for the long-term treatment of radioactive graphite waste which have ranged from sea dumping through incineration to land-based disposal, sometimes preceded by a variable period of 'safe storage' within the original reactor containment. Nuclear regulators are challenging the proposed length of 'safe storage' on the basis that essential knowledge may be lost. More recently, political constraints have further complicated the issue by eliminating disposal at sea and imposing a 'near-zero release' philosophy, while public opinion is opposed to land-based disposal and has induced a continual drive towards minimizing radioactivity release to the environment from disposal. This paper proposes that, despite various international agreements, it is time to review technically all options for disposal of irradiated graphite waste as a framework for the eventual decision-making process. It is recognized that the socio-economic and political pressures are high and therefore, given that all currently identified options satisfy the present safety limits, the need to minimize the objective risk is shown to be a minor need in comparison to the public's want of demonstrable control, responsiveness and ability to reverse/change the disposal options in the future. Further, it is shown that the eventual decision-making process for a post-dismantling option for graphite waste must optimize the beneficial attributes of subjective risk experienced by the general public. In addition, in advocating and preferred option to the general public, it is recommended that the industry should communicate at a level commensurate with the public understanding and initiate a process of facilitation which enables the public to arrive at their own solution and constituting a social exchange. Otherwise it is concluded that if the indecision over disposal options is allowed to continue then, by default, graphite will remain in long-term supervised storage. (author)

Nuclear graphite for high temperature reactors

International Nuclear Information System (INIS)

Marsden, B.J.

2001-01-01

The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

Irradiation creep performance of graphite relevant for pebble bed HTRs

International Nuclear Information System (INIS)

Kleist, G.; O'Connor, M.F.

1980-01-01

Irradiation - induced creep in the core reflector component graphite of high temperature reactors is of primary importance to the core designer since it provides a mechanism for the relief of internal stresses arising from differential Wigner shrinkage and thermal expansion. The experimental determination of the extent of this creep for conditions relevant to the reactor is thus imperative

Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

International Nuclear Information System (INIS)

Le-Guillou, Mael

2014-01-01

In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

Behavior of LASL-made graphite, ZrC, and ZrC-containing coated particles in irradiation tests HT-28 and HT-29

International Nuclear Information System (INIS)

Reiswig, R.D.; Wagner, P.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

1976-01-01

Three types of materials, extruded graphite, hot-pressed ZrC, and particles with ZrC coatings, were irradiated in ORNL High Fluence Isotope Reactor Irradiation tests HT-28 and HT-29. The ZrC seemed unaffected. The graphite changed in dimensions, x-ray diffraction parameters, and thermal conductivity. The four types of coated particles tested all resisted the irradiation well, except one set of particles with double-graded C-ZrC-C coats. Overall, the results were considered encouraging for use of ZrC and extruded graphite fuel matrices. 16 fig

Simulating Neutron Radiation Damage of Graphite by In-situ Electron Irradiation

International Nuclear Information System (INIS)

Mironov, Brindusa E; Freeman, H M; Brydson, R M D; Westwood, A V K; Scott, A J

2014-01-01

Radiation damage in nuclear grade graphite has been investigated using transmission electron microscopy (TEM) and electron energy loss spectroscopy (EELS). Changes in the structure on the atomic scale and chemical bonding, and the relationship between each were of particular interest. TEM was used to study damage in nuclear grade graphite on the atomic scale following 1.92×10 8 electrons nm −2 of electron beam exposure. During these experiments EELS spectra were also collected periodically to record changes in chemical bonding and structural disorder, by analysing the changes of the carbon K-edge. Image analysis software from the 'PyroMaN' research group provides further information, based on (002) fringe analysis. The software was applied to the micrographs of electron irradiated virgin 'Pile Grade A' (PGA) graphite to quantify the extent of damage from electron beam exposure

Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

International Nuclear Information System (INIS)

Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

1999-01-01

This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

Erosion of CFC, pyrolytic and boronated graphite under short pulsed laser irradiation

International Nuclear Information System (INIS)

Kraaij, G.J.; Bakker, J.; Stad, R.C.L. van der

1992-07-01

The effect of short pulsed laser irradiation of '0/3' ms and up to 10 MJ/m 2 on different types of carbon base materials is described. These materials are investigated as candidate protection materials for the Plasma Facing Components of NET/ITER. These materials are: carbon fibre composite graphite, pyrolytic graphite and boronated graphite. The volume of the laser induced craters was measured with an optical topographic scanner, and these data are evaluated with a simple model for the erosion. As a results, the enthalpy of ablation is estimated as 30±3 MJ/kg. A comparison is made with finite element numerical calculations, and the effect of lateral heat transfer is estimated using an analytical model. (author). 8 refs., 23 figs., 4 tabs

Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

OpenAIRE

Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

2015-01-01

Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

Phonon scattering in graphite

International Nuclear Information System (INIS)

Wagner, P.

1976-04-01

Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

Innovative approaches to the Management of Irradiated Nuclear Graphite Wastes: Addressing the Challenges through International Collaboration with Project 'GRAPA'

International Nuclear Information System (INIS)

Wickham, A.J.; Ojovan, M.; O'Sullivan, P.; )

2017-01-01

There exists more than 250.000 tonnes of irradiated (and therefore radioactive) nuclear graphite (i-graphite) in the world, primarily as a result of the development of graphite-moderated power-reactor systems, initially for defence and subsequently for commercial purposes. Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous), the latter producing a continuous stream of so-called 'operational waste'. The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the European Union and the International Atomic Energy Agency for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. An IAEA collaborative research programme on treatment options has recently been completed, and an active group of international specialists in this area has now been established as part of the IAEA International Decommissioning Network under the envelope of Project 'GRAPA' (Irradiated Graphite Processing Approaches), which includes representatives from Belgium, China, France, Germany, India, Italy, Lithuania, Rep. of Korea, Romania, Spain, Switzerland, Ukraine and the Russian Federation with direct responsibilities for various parts of the decommissioning and graphite-disposal process in a variety of reactor designs. Interest has also been expressed by colleagues from Sweden and Japan. Work is in progress on a number of topic areas where weaknesses in the

Simultaneous heating and compression of irradiated graphite during synchrotron microtomographic imaging

Science.gov (United States)

Bodey, A. J.; Mileeva, Z.; Lowe, T.; Williamson-Brown, E.; Eastwood, D. S.; Simpson, C.; Titarenko, V.; Jones, A. N.; Rau, C.; Mummery, P. M.

2017-06-01

Nuclear graphite is used as a neutron moderator in fission power stations. To investigate the microstructural changes that occur during such use, it has been studied for the first time by X-ray microtomography with in situ heating and compression. This experiment was the first to involve simultaneous heating and mechanical loading of radioactive samples at Diamond Light Source, and represented the first study of radioactive materials at the Diamond-Manchester Imaging Branchline I13-2. Engineering methods and safety protocols were developed to ensure the safe containment of irradiated graphite as it was simultaneously compressed to 450N in a Deben 10kN Open-Frame Rig and heated to 300°C with dual focused infrared lamps. Central to safe containment was a double containment vessel which prevented escape of airborne particulates while enabling compression via a moveable ram and the transmission of infrared light to the sample. Temperature measurements were made in situ via thermocouple readout. During heating and compression, samples were simultaneously rotated and imaged with polychromatic X-rays. The resulting microtomograms are being studied via digital volume correlation to provide insights into how thermal expansion coefficients and microstructure are affected by irradiation history, load and heat. Such information will be key to improving the accuracy of graphite degradation models which inform safety margins at power stations.

Effect of various dopant elements on primary graphite growth

International Nuclear Information System (INIS)

Valle, N; Theuwissen, K; Lacaze, J; Sertucha, J

2012-01-01

Five spheroidal graphite cast irons were investigated, a usual ferritic grade and four pearlitic alloys containing Cu and doped with Sb, Sn and Ti. These alloys were remelted in a graphite crucible, leading to volatilization of the magnesium added for spheroidization and to carbon saturation of the liquid. The alloys were then cooled down and maintained at a temperature above the eutectic temperature. During this step, primary graphite could develop showing various features depending on the doping elements added. The largest effects were that of Ti which greatly reduces graphite nucleation and growth, and that of Sb which leads to rounded agglomerates instead of lamellar graphite. The samples have been investigated with secondary ion mass spectrometry to enlighten distribution of elements in primary graphite. SIMS analysis showed almost even distribution of elements, including Mg and Al (from the inoculant) in the ferritic grade, while uneven distribution was evident in all doped alloys. Investigations are going on to clarify if the uneven distribution is associated with structural defects in the graphite precipitates.

AGC-2 Graphite Preirradiation Data Package

Energy Technology Data Exchange (ETDEWEB)

David Swank; Joseph Lord; David Rohrbaugh; William Windes

2012-10-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors

International Nuclear Information System (INIS)

Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Kilda, Raimondas

2010-01-01

Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit I reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases. (orig.)

An investigation of the electron irradiation of graphite in a helium atmosphere using a modified electron microscope

International Nuclear Information System (INIS)

Burden, A.P.; Hutchison, J.L.

1997-01-01

The behaviour of graphite particles immersed in helium gas and irradiated with an electron-beam has been investigated. Because this treatment was performed in a modified high resolution transmission electron microscope, the rapid morphological and microstructural changes that occurred could be directly observed. The results have implications for future controlled environment microscopy of carbonaceous materials and the characterisation of such microscopes. It is also shown that the processes can provide insight into ion-irradiation induced damage of graphite and the mechanism of fullerene generation. (Author)

Characterization of fresh and irradiated domestic nuclear graphite; Karakterizacija neozracenog i ozracenog domaceg nuklearnog grafita

Energy Technology Data Exchange (ETDEWEB)

Marinkovic, S; Suznjevic, C; Bogdanovic, R; Gasic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

This report shows results of testing the quality of domestic impregnated graphite IGSP-05, and irradiated domestic graphite IGSP-01 as well as the new methos of characterization based on graphite oxidation by liquid agent. Systematic measurement of domestic impregnated graphite enabled conclusions related to its quality and further improvement. Domestic graphite is relatively well graphitized and its properties are approaching standard nuclear graphite, although it still shows some deficiencies. Important deficiencies are significant inhomogeneity and low density. The applied impregnation procedure did not improve significantly the quality of graphite, probably because the material which was impregnated had fine pores. To avoid this porosity it would be necessary to use material with higher granulation. Soot which was present in some blocks probably worsened the quality of graphite and caused dispersion of the obtained results. First tests of irradiated domestic graphite IGSP-01 showed that its behaviour does not differ from standard nuclear graphite in case of low doses. It is necessary to test its properties in case of higher neutron doses before drawing final conclusions. The new method of graphite oxidation by the N{sub 2}SO{sub 4} - Ag{sub 2}Cr{sub 2}O{sub 7} mixture which is highly sensitive on the existence of structural defects is based on detecting the oxidation rate of graphite by measuring the pressure of released CO{sub 2}. Application of the method for testing the domestic and American graphite showed that irradiation caused drastic changes of oxidation rates and similar behaviour of both graphite types. U ovom izvestaju su prikazani rezultati ispitivanja kvaliteta domaceg impregnisanog grafita IGSP-05, rezultati ispitivanja ozracenog domaceg grafita IGSP-01 i opisana je nova uvedena metoda karakterizacije zasnovana na oksidaciji grafita tecnim agensom. Sistematsko merenje osobina domaceg impregnisanog grafita je omogucilo donosenje zakljucaka o

Special graphites

International Nuclear Information System (INIS)

Leveque, P.

1964-01-01

A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

AGC-2 Graphite Preirradiation Data Analysis Report

Energy Technology Data Exchange (ETDEWEB)

William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

2013-08-01

This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

Effects of high temperature neutron irradiation on the physical, chemical and mechanical properties of fine-grained isotropic graphite

International Nuclear Information System (INIS)

Matsuo, H.; Nomura, S.; Imai, H.; Oku, T.; Eto, M.

1987-01-01

Effects of neutron irradiation on the dimensional change, coefficient of thermal expansion(CTE), thermal conductivity, corrosion rate, Young's modulus and strengths were studied for the candidate graphite material IG-110 of the experimental very high temperature gas-cooled reactor(VHTR) after irradiation at 585 - 1273 deg C to neutron fluences of up to about 3 x 10 25 n/m 2 (E > 29 fJ) in the JMTR and JRR-2, and to about 7 x 10 25 n/m 2 (E > 29 fJ) in the HFR. The results were compared with the irradiation behaviors of other graphites. Dimensional shrinkage was observed in the whole irradiation temperature range, showing lower value than 2 %. The shrinkage rate showed the minimum in the irradiation temperature of around 850 deg C, followed by the increase for the samples irradiated at higher temperatures. The dimensional stability of the material was clarified to be almost the same with that of H451 graphite. The CTE, thermal resistivity and Young's modulus increased in the early stage of irradiation and then only the CTE decreased while the thermal resistivity and Young's modulus levelled off with further irradiation. The neutron fluence showing the maximum CTE shifted to the lower fluence with increasing irradiation temperature. The increases of both thermal resistivity and Young's modulus were remarkable for the samples irradiated at lower temperatures. Compressive and bending strengths measured at room temperature increased after irradiation as well. The corrosion rate with water-vapor of 0.65 % in helium at high temperatures decreased owing to irradiation and the reduction was independent of irradiation temperature and neutron fluence. The activation energy for the reaction was estimated to be the same before and after irradiation. (author)

Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

International Nuclear Information System (INIS)

Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

2005-01-01

G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

Nuclear graphite ageing and turnaround

International Nuclear Information System (INIS)

Marsden, B.J.; Hall, G.N.; Smart, J.

2001-01-01

Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

Evaluation of thermal shock strengths for graphite materials using a laser irradiation method

International Nuclear Information System (INIS)

Kim, Jae Hoon; Lee, Young Shin; Kim, Duck Hoi; Park, No Seok; Suh, Jeong; Kim, Jeng O.; Il Moon, Soon

2004-01-01

Thermal shock is a physical phenomenon that occurs during the exposure to rapidly high temperature and pressure changes or during quenching of a material. The rocket nozzle throat is exposed to combustion gas of high temperature. Therefore, it is important to select suitable materials having the appropriate thermal shock resistance and to evaluate these materials for rocket nozzle design. The material of this study is ATJ graphite, which is the candidate material for rocket nozzle throat. This study presents an experimental method to evaluate the thermal shock resistance and thermal shock fracture toughness of ATJ graphite using laser irradiation. In particular, thermal shock resistance tests are conducted with changes of specimen thickness, with laser source irradiated at the center of the specimen. Temperature distributions on the specimen surface are detected using type K and C thermocouples. Scanning electron microscope (SEM) is used to observe the thermal cracks on specimen surface

AGC-3 Graphite Preirradiation Data Analysis Report

Energy Technology Data Exchange (ETDEWEB)

William Windes; David Swank; David Rohrbaugh; Joseph Lord

2013-09-01

This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen’s 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to

Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

International Nuclear Information System (INIS)

Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Tobita, Tsutomu; Kikuchi, Teruo; Kurobane, Shiro; Adachi, Mamoru; Fukuda, Kousaku

1988-06-01

Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 10 24 n/m 2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

Direct growth of self-crystallized graphene and graphite nanoballs with Ni vapor-assisted growth: from controllable growth to material characterization.

Science.gov (United States)

Yen, Wen-Chun; Chen, Yu-Ze; Yeh, Chao-Hui; He, Jr-Hau; Chiu, Po-Wen; Chueh, Yu-Lun

2014-05-09

A directly self-crystallized graphene layer with transfer-free process on arbitrary insulator by Ni vapor-assisted growth at growth temperatures between 950 to 1100 °C via conventional chemical vapor deposition (CVD) system was developed and demonstrated. Domain sizes of graphene were confirmed by Raman spectra from ~12 nm at growth temperature of 1000 °C to ~32 nm at growth temperature of 1100 °C, respectively. Furthermore, the thickness of the graphene is controllable, depending on deposition time and growth temperature. By increasing growth pressure, the growth of graphite nano-balls was preferred rather than graphene growth. The detailed formation mechanisms of graphene and graphite nanoballs were proposed and investigated in detail. Optical and electrical properties of graphene layer were measured. The direct growth of the carbon-based materials with free of the transfer process provides a promising application at nanoelectronics.

Nanoscale transformation of sp2 to sp3 of graphite by slow highly charged ion irradiation

International Nuclear Information System (INIS)

Meguro, T.; Hida, A.; Koguchi, Y.; Miyamoto, S.; Yamamoto, Y.; Takai, H.; Maeda, K.; Aoyagi, Y.

2003-01-01

Nanoscale transformation of electronic states by highly charged ion (HCI) impact on graphite surfaces is described. The high potential energy of slow HCI, which induces multiple emission of electrons from the surface, provides a strong modification of the electronic states of the local area upon graphite surfaces. The HCI impact and the subsequent surface treatment either by electron injection from a scanning tunneling microscopy tip or by He-Cd laser irradiation induce a localized transition from sp 2 to sp 3 hybridization in graphite, resulting in the formation of nanoscale diamond-like structures (nanodiamond) at the impact region. From Raman spectroscopic measurements on sp 2 related peaks, it is found that the HCI irradiation creates vacancy complexes in contrast to ions having a lower charge state, which generate single vacancies. It is of interest that a single impact of HCI creates one nanodiamond structure, suggesting potential applications of HCI in nanoscale material processing

Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite

International Nuclear Information System (INIS)

Bocquet, M.; Micaud, G.

1962-01-01

An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient ΔΘ/ Δx in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. Δx/ΔΘ, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [fr

A discussion of possible mechanisms affecting fission product transport in irradiated and unirradiated nuclear grade graphite

International Nuclear Information System (INIS)

Firth, M.J.

1977-09-01

137 Cs, 85 Sr, and sup(110m)Ag adsorption experiments were conducted on three graphite powders with differing amounts of specific basal and edge surface areas. No direct proportionality was found between the specific amounts of the isotopes adsorbed and either of the surface characteristics. There appears to be some correlation with the specific basal surface area despite the fact that each isotope behaves differently. Factors that might influence the adsorption behaviour of Cs and Ag during reactor irradiation and heat treatment of nuclear grade graphites are discussed. These include the form of Cs with the graphite surface. A model based on Cs adsorption at vacancy clusters is used to analyse adsorption experiments. A possible explanation for the behaviour of Ag through the migration of graphite impurities from the bulk of the graphite to the pore surface is also discussed. (author)

Direct growth of self-crystallized graphene and graphite nanoballs with Ni vapor-assisted growth: From controllable growth to material characterization

Science.gov (United States)

Yen, Wen-Chun; Chen, Yu-Ze; Yeh, Chao-Hui; He, Jr-Hau; Chiu, Po-Wen; Chueh, Yu-Lun

2014-01-01

A directly self-crystallized graphene layer with transfer-free process on arbitrary insulator by Ni vapor-assisted growth at growth temperatures between 950 to 1100°C via conventional chemical vapor deposition (CVD) system was developed and demonstrated. Domain sizes of graphene were confirmed by Raman spectra from ~12 nm at growth temperature of 1000°C to ~32 nm at growth temperature of 1100°C, respectively. Furthermore, the thickness of the graphene is controllable, depending on deposition time and growth temperature. By increasing growth pressure, the growth of graphite nano-balls was preferred rather than graphene growth. The detailed formation mechanisms of graphene and graphite nanoballs were proposed and investigated in detail. Optical and electrical properties of graphene layer were measured. The direct growth of the carbon-based materials with free of the transfer process provides a promising application at nanoelectronics. PMID:24810224

Study of graphitic microstructure formation in diamond bulk by pulsed Bessel beam laser writing

Science.gov (United States)

Kumar, S.; Sotillo, B.; Chiappini, A.; Ramponi, R.; Di Trapani, P.; Eaton, S. M.; Jedrkiewicz, O.

2017-11-01

The advantages of using Bessel beams for the generation of graphitic structures in diamond bulk are presented. We show that by irradiating the sample with a pulsed Bessel beam whose non-diffracting zone is of the same order of the sample thickness, it is possible to produce without any sample translation straight graphitic through-microstructures, whose size depends on the input pulse energy. The microstructure growth is investigated as a function of the number of irradiating pulses, and the femtosecond and picosecond regimes are contrasted.

Graphite selection for the FMIT test cell

International Nuclear Information System (INIS)

Morgan, W.C.

1982-06-01

This document provides the basis for procuring a grade of graphite, at minimum cost, having minimum dimensional changes at low irradiation temperatures (nominal range 90 to 140 0 C). In light of those constraints, the author concludes that the most feasible approach is to attempt to reproduce a grade of graphite (TSGBF) which has exhibited a high degree of dimensional stability during low-temperature irradiations and on which irradiation-induced changes in other physical properties have been measured. The effects of differences in raw materials, especially coke morphology, and processing conditions, primarily graphitization temperture are briefly reviewed in terms of the practicality of producing a new grade of graphite with physical properties and irradiation-induced changes which would be very similar to those of TSGBF graphite. The production history and physical properties of TSGBF are also reviewed; no attempt is made, to project changes in dimensions or physical properties under the projected irradiation conditions

Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

Science.gov (United States)

Remeikis, Vidmantas; Lagzdina, Elena; Garbaras, Andrius; Gudelis, Arūnas; Garankin, Jevgenij; Plukienė, Rita; Juodis, Laurynas; Duškesas, Grigorijus; Lingis, Danielius; Abdulajev, Vladimir; Plukis, Artūras

2018-01-01

14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

Directory of Open Access Journals (Sweden)

Vidmantas Remeikis

Full Text Available 14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

Thermal desorption spectroscopy of pyrolytic graphite cleavage faces after keV deuterium irradiation at 330-1000 K

International Nuclear Information System (INIS)

Gotoh, Y.; Yamaki, T.; Tokiguchi, K.

1992-01-01

Thermal desorption spectroscopy (TDS) measurements were made on D 2 and CD 4 from surface layers of pyrolytic graphite cleavage faces after 3 keV D + 3 irradiation to 1.5 x 10 18 D/cm 2 at irradiation temperatures from 330 to 1000 K. Thermal desorption of both D 2 and CD 4 was observed to rise simultaneously at around 700 K. The D 2 peak was found at T m = 900-1000 K, while the CD 4 peak appeared at a lower temperature, 800-840 K. The T m for the D 2 TDS increased, while that for the CD 4 decreased with increasing irradiation temperature. These results obviously indicate that the D 2 desorption is detrapping/recombination limited, while the CD 4 desorption is most likely to be diffusion limited. The amount of thermally desorbed D 2 after the D + irradiation was observed to monotonously decrease as the irradiation temperature was increased from 330 to 1000 K. These tendencies agreed with previous results for the irradiation temperature dependencies of both C1s chemical shift (XPS) and the interlayer spacing, d 002 (HRTEM), on the graphite basal face. (orig.)

Technical specifications (replaces note T.62). Irradiation of graphite at ambient temperature, Note T. 76; Specification technique, (Annule et remplace la note T. 62), Irradiation de graphite a temperature ambiante, Note T. 76

Energy Technology Data Exchange (ETDEWEB)

Reseau, R A [Services des grandes piles experimentales, Section ' Physique et Experimentation, Saclay (France)

1962-12-15

The objective is to study the effects of fast neutron irradiation of different graphite samples. The irradiation conditions should be as follows: integral fast neutron flux should be higher than 10{sup 20} neutrons/cm{sup 2}, the reactor should operate at steady state for 15 days, the temperature od samples should not be higher than 100 deg C, preferably 80 deg C. Note T. 62 which is replaced by this Note is attached.

Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

International Nuclear Information System (INIS)

Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

1995-01-01

This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

Energy Technology Data Exchange (ETDEWEB)

Ammigan, K. [Fermilab; Hurh, P. [Fermilab; Sidorov, V. [Fermilab; Zwaska, R. [Fermilab; Asner, D. M. [PNL, Richland; Casella, Casella,A.M [PNL, Richland; Edwards, D. J. [PNL, Richland; Schemer-Kohrn, A. L. [PNL, Richland; Senor, D. J. [PNL, Richland

2017-02-10

The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

A systematic study of acoustic emission from nuclear graphites

International Nuclear Information System (INIS)

Neighbour, G.B.; McEnaney, B.

1996-01-01

Acoustic emission (AE) monitoring has been identified as a possible method to determine internal stresses in nuclear graphites using the Kaiser effect, i.e., on stressing a graphite that has been subject to a prior stress, the onset of AE occurs at the previous peak stress. For three nuclear graphites (PGA, IM1-24 and VNEC), AE was monitored during both monotonic and cyclic loading to failure in tensile, compressive and flexural test modes. For unirradiated graphites, the Kaiser effect was not found in cyclic loading, but a Felicity effect was observed, i.e., the onset of AE occurred below the previously applied peak stress. The Felicity effect was attributed to time-dependent relaxation and recovery processes and was characterized using a new parameter, the Recovery ratio. It was shown that AE can be used to monitor creep strain and creep recovery in graphites at zero load. The AE-time responses from these experiments were fitted to equations similar to those used for creep strain-time at elevated temperatures. The number of AE counts from irradiated graphites were greater than those from unirradiated graphites, subject to similar stresses, due to increases in porosity caused by radiolytic oxidation. A Felicity effect was also observed on cyclic loading of irradiated graphites, but no evidence for a Kaiser effect was found for irradiated graphites loaded monotonically to failure. Thus internal stresses in irradiated graphites could not be measured using AE. This was attributed to relaxation and recovery processes that occur between removing the irradiated graphite from the reactor and AE testing. This work indicated that AE monitoring is not a suitable technique for measuring internal stresses in irradiated graphite. (author). 19 refs, 6 figs, 6 tabs

Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR.

Science.gov (United States)

Zhang, Z T; Xu, C; Dmytriieva, D; Molatta, S; Wosnitza, J; Wang, Y T; Helm, M; Zhou, Shengqiang; Kühne, H

2017-10-20

We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by 13 C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the 13 C nuclear spin-lattice relaxation rate [Formula: see text] by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of [Formula: see text] below about 10 K can well be described by a thermally activated form, [Formula: see text], yielding a singular Zeeman energy of ([Formula: see text]) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments.

Graphite irradiated by swift heavy ions under grazing incidence

CERN Document Server

Liu, J; Müller, C; Neumann, R

2002-01-01

Highly oriented pyrolytic graphite is irradiated with various heavy projectiles (Ne, Ni, Zn, Xe and U) in the MeV to GeV energy range under different oblique angles of incidence. Using scanning tunneling microscopy, the impact zones are imaged as hillocks protruding from the surface. The diameter of surface-grazing tracks varies between 3 nm (Ne) and 6 nm (U), which is about twice as large as under normal beam incidence. Exclusively for U and Xe projectiles, grazing tracks exhibit long comet-like tails consisting of successive little bumps indicating that the damage along the ion path is discontinuous even for highest electronic stopping powers.

Understanding the Growth Mechanism of GaN Epitaxial Layers on Mechanically Exfoliated Graphite.

Science.gov (United States)

Li, Tianbao; Liu, Chenyang; Zhang, Zhe; Yu, Bin; Dong, Hailiang; Jia, Wei; Jia, Zhigang; Yu, Chunyan; Gan, Lin; Xu, Bingshe; Jiang, Haiwei

2018-04-27

The growth mechanism of GaN epitaxial layers on mechanically exfoliated graphite is explained in detail based on classic nucleation theory. The number of defects on the graphite surface can be increased via O-plasma treatment, leading to increased nucleation density on the graphite surface. The addition of elemental Al can effectively improve the nucleation rate, which can promote the formation of dense nucleation layers and the lateral growth of GaN epitaxial layers. The surface morphologies of the nucleation layers, annealed layers and epitaxial layers were characterized by field-emission scanning electron microscopy, where the evolution of the surface morphology coincided with a 3D-to-2D growth mechanism. High-resolution transmission electron microscopy was used to characterize the microstructure of GaN. Fast Fourier transform diffraction patterns showed that cubic phase (zinc-blend structure) GaN grains were obtained using conventional GaN nucleation layers, while the hexagonal phase (wurtzite structure) GaN films were formed using AlGaN nucleation layers. Our work opens new avenues for using highly oriented pyrolytic graphite as a substrate to fabricate transferable optoelectronic devices.

AGC-2 Irradiation Report

Energy Technology Data Exchange (ETDEWEB)

Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

Graphite for high-temperature reactors

International Nuclear Information System (INIS)

Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

1976-01-01

The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

Adsorption of dysprosium on the graphite (0001) surface: Nucleation and growth at 300 K

International Nuclear Information System (INIS)

Kwolek, Emma J.; Lii-Rosales, Ann; Lei, Huaping; Wang, Cai-Zhuang; Tringides, Michael C.; Evans, James W.; Wallingford, Mark; Zhou, Yinghui; Thiel, Patricia A.

2016-01-01

We have studied nucleation and growth of Dy islands on the basal plane of graphite at 300 K using scanning tunneling microscopy, density functional theory (DFT) in a form that includes van der Waals interactions, and analytic theory. The interaction of atomic Dy with graphite is strong, while the diffusion barrier is small. Experiment shows that at 300 K, the density of nucleated islands is close to the value predicted for homogeneous nucleation, using critical nucleus size of 1 and the DFT-derived diffusion barrier. Homogeneous nucleation is also supported by the monomodal shape of the island size distributions. Comparison with the published island density of Dy on graphene shows that the value is about two orders of magnitude smaller on graphite, which can be attributed to more effective charge screening in graphite. The base of each island is 3 atomic layers high and atomically ordered, forming a coincidence lattice with the graphite. Islands resist coalescence, probably due to multiple rotational orientations associated with the coincidence lattice. Upper levels grow as discernible single-atom layers. Analysis of the level populations reveals significant downward interlayer transport, which facilitates growth of the base. This island shape is metastable, since more compact three-dimensional islands form at elevated growth temperature.

Adsorption of dysprosium on the graphite (0001) surface: Nucleation and growth at 300 K

Energy Technology Data Exchange (ETDEWEB)

Kwolek, Emma J.; Lii-Rosales, Ann [The Ames Laboratory, Ames, Iowa 50011 (United States); Department of Chemistry, Iowa State University, Ames, Iowa 50011 (United States); Lei, Huaping; Wang, Cai-Zhuang; Tringides, Michael C.; Evans, James W. [The Ames Laboratory, Ames, Iowa 50011 (United States); Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Wallingford, Mark; Zhou, Yinghui [The Ames Laboratory, Ames, Iowa 50011 (United States); Thiel, Patricia A., E-mail: pthiel@iastate.edu [The Ames Laboratory, Ames, Iowa 50011 (United States); Department of Chemistry, Iowa State University, Ames, Iowa 50011 (United States); Department of Materials Science and Engineering, Iowa State University, Ames, Iowa 50011 (United States)

2016-12-07

We have studied nucleation and growth of Dy islands on the basal plane of graphite at 300 K using scanning tunneling microscopy, density functional theory (DFT) in a form that includes van der Waals interactions, and analytic theory. The interaction of atomic Dy with graphite is strong, while the diffusion barrier is small. Experiment shows that at 300 K, the density of nucleated islands is close to the value predicted for homogeneous nucleation, using critical nucleus size of 1 and the DFT-derived diffusion barrier. Homogeneous nucleation is also supported by the monomodal shape of the island size distributions. Comparison with the published island density of Dy on graphene shows that the value is about two orders of magnitude smaller on graphite, which can be attributed to more effective charge screening in graphite. The base of each island is 3 atomic layers high and atomically ordered, forming a coincidence lattice with the graphite. Islands resist coalescence, probably due to multiple rotational orientations associated with the coincidence lattice. Upper levels grow as discernible single-atom layers. Analysis of the level populations reveals significant downward interlayer transport, which facilitates growth of the base. This island shape is metastable, since more compact three-dimensional islands form at elevated growth temperature.

Irradiation-induced creep in graphite: a review

International Nuclear Information System (INIS)

Price, R.J.

1981-08-01

Data on irradiation-induced creep in graphite published since 1972 are reviewed. Sources include restrained shrinkage tests conducted at Petten, the Netherlands, tensile creep experiments with continuous strain registration at Petten and Grenoble, France, and controlled load tests with out-of-reactor strain measurement performed at Oak Ridge National Laboratory, Petten, and in the United Kingdom. The data provide reasonable confirmation of the linear viscoelastic creep model with a recoverable transient strain component followed by a steady-state strain component, except that the steady-state creep coefficient must be treated as a function of neutron fluence and is higher for tensile loading than for compressive loading. The total transient creep strain is approximately equal to the preceding elastic strain. No temperature dependence of the transient creep parameters has been demonstrated. The initial steady-state creep coefficient is inversely proportional to the unirradiated Young modulus

Comparison of 3 MeV C+ Ion-Irradiation Effects between The Nuclear Graphites made of Pitch and Petroleum Cokes

International Nuclear Information System (INIS)

Se-Hwan, Chi; Gen-Chan, Kim; Jong-Hwa, Chang

2006-01-01

Currently, all the commercially available nuclear graphite grades are being made from two different cokes, i.e., petroleum coke or coal-tar pitch coke, and a coal-tar pitch binder. Of these, since the coke composes most of the graphite volume, i.e., > 70 %, it is understood that a physical, chemical, thermal, and mechanical property as well as an irradiation-induced property change will be strongly dependent on the type of coke. To obtain first-hand information on the effects of the coke type, i.e., petroleum or pitch, on the irradiation sensitivity of graphite, specimens made of IG-110 of petroleum coke and IG-430 of pitch coke were irradiated up to ∼ 19 dpa by 3 MeV C + at room temperature, and the irradiation-induced changes in the hardness, Young's modulus, Raman spectrum, and oxidation properties were characterized. Results of the TEM show that the size and density of the Mrozowski cracks appeared to be far larger and higher in the IG-110 than the IG-430. Results of the hardness test revealed a slightly higher increase in the IG-430 than the IG-110 by around 10 dpa, and the Raman spectrum measurement showed a higher (FWHM) D /(FWHM) G value for IG-430 for 0.02 ∼ 0.25 dpa. Both the hardness and Raman measurement may imply a higher irradiation sensitivity of the IG-430 than the IG-110. Results of the Young's modulus measurements showed a large data scattering, which prevented us from estimating the differences between the grades. Oxidation experiments using a TG-DTA under a flow of dry air/He = 2.5 % (flow rate: 40 CC/min) at 750 and 1000 deg C show that the IG-110 of the petroleum coke exhibits a far higher oxidation rate than the IG-430. The discrepancy between the oxidation rate of the two grades increased with an increase in the oxidation temperature and the dose. Oxidized surface pore area was larger for IG-110. Judging from the results obtained from the present experimental conditions, the irradiation sensitivity appeared to be dependent on the degree

Study on structural recovery of graphite irradiated with swift heavy ions at high temperature

Energy Technology Data Exchange (ETDEWEB)

Pellemoine, F., E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Avilov, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Bender, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Ewing, R.C. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Fernandes, S. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Lang, M. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996-2300 (United States); Li, W.X. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Mittig, W. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Schein, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Severin, D. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Tomut, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Laboratory of Magnetism and Superconductivity, National Institute for Materials Physics NIMP, Bucharest (Romania); Trautmann, C. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Dept. of Materials Science, Technische Universität Darmstadt, Darmstadt (Germany); and others

2015-12-15

Thin graphite foils bombarded with an intense high-energy (8.6 MeV/u) gold beam reaching fluences up to 1 × 10{sup 15} ions/cm{sup 2} lead to swelling and electrical resistivity changes. As shown earlier, these effects are diminished with increasing irradiation temperature. The work reported here extends the investigation of beam induced changes of these samples by structural analysis using synchrotron X-ray diffraction and transmission electron microscope. A nearly complete recovery from swelling at irradiation temperatures above about 1500 °C is identified.

Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR.

Science.gov (United States)

Zhang, Zhi Tao; Xu, C; Dmytriieva, Daryna; Molatta, Sebastian; Wosnitza, J; Wang, Y T; Helm, Manfred; Zhou, Shengqiang; Kuehne, Hannes

2017-09-18

We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by $^{13}$C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the $^{13}$C nuclear spin-lattice relaxation rate $1/T_{1}$ by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of $1/T_{1}$ below about 10 K can well be described by a thermally activated form, $1/T_{1}\\propto\\exp(-\\Delta/k_{B}T)$, yielding a singular Zeeman energy of ($0.41\\pm0.01$) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments. © 2017 IOP Publishing Ltd.

Study by electronic microscopy of corrosion features of graphite after hot oxidation (air, 620 C)

International Nuclear Information System (INIS)

Jodon de Villeroche, Suzanne

1968-01-01

The author reports the study of corrosion features of graphite after hot oxidation in the air at 620 C. It is based on observations made by electronic microscopy. This study comes after another one dedicated to oxidation features obtained by hot corrosion of natural graphite, and aims at comparing pyrolytic graphite before and after irradiation in an atomic pile, and at performing tests on a graphite processed with ozone. After a recall of generalities about natural graphite and of some issues related to hot corrosion of natural graphite, the author presents some characteristics and features of irradiated and non-irradiated pyrolytic graphite. He reports the study of the oxidation of samples of pyrolytic graphite: production of thin lamellae, production of glaze-carbon replicates, oxidation of irradiated and of non-irradiated graphite, healing of irradiation defects, and oxidation of ozone-processed natural graphite [fr

Influence of irradiation on high-strength graphites

International Nuclear Information System (INIS)

Virgil'ev, Yu.S.; Grebennik, V.N.; Kalyagina, I.P.

1989-01-01

To ensure efficiency of the graphite elements of the construction of the masonry of reactors, the graphite must possess high radiation stability, strength, and heat resistance. In this connection, the physical properties of graphites based on uncalcined petroleum coke with a binder - high-temperature hard coal pitch - the amount of which reaches 40% are considered in this paper

The Impact of Alkaliphilic Biofilm Formation on the Release and Retention of Carbon Isotopes from Nuclear Reactor Graphite.

Science.gov (United States)

Rout, S P; Payne, L; Walker, S; Scott, T; Heard, P; Eccles, H; Bond, G; Shah, P; Bills, P; Jackson, B R; Boxall, S A; Laws, A P; Charles, C; Williams, S J; Humphreys, P N

2018-03-13

14 C is an important consideration within safety assessments for proposed geological disposal facilities for radioactive wastes, since it is capable of re-entering the biosphere through the generation of 14 C bearing gases. The irradiation of graphite moderators in the UK gas-cooled nuclear power stations has led to the generation of a significant volume of 14 C-containing intermediate level wastes. Some of this 14 C is present as a carbonaceous deposit on channel wall surfaces. Within this study, the potential of biofilm growth upon irradiated and 13 C doped graphite at alkaline pH was investigated. Complex biofilms were established on both active and simulant samples. High throughput sequencing showed the biofilms to be dominated by Alcaligenes sp at pH 9.5 and Dietzia sp at pH 11.0. Surface characterisation revealed that the biofilms were limited to growth upon the graphite surface with no penetration of the deeper porosity. Biofilm formation resulted in the generation of a low porosity surface layer without the removal or modification of the surface deposits or the release of the associated 14 C/ 13 C. Our results indicated that biofilm formation upon irradiated graphite is likely to occur at the pH values studied, without any additional release of the associated 14 C.

Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

Energy Technology Data Exchange (ETDEWEB)

Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

2013-07-01

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

Nuclear graphite waste management. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2001-05-01

The purpose of the seminar was to bring together the specialists dealing with various aspects of radioactive graphite waste management to exchange and review information on the decommissioning, characterisation, processing and disposal of irradiated graphite from reactor cores and other graphite waste associated with reactor operation. The seminar covered radioactive graphite characterisation, the effect of irradiation on graphite components, Wigner energy, radioactive graphite waste treatment, conditioning, interim storage and long term disposal options. Individual papers presented at the seminar were indexed separately

Contribution to the study of internal friction in graphites; Contribution a l'etude du frottement interieur des graphites

Energy Technology Data Exchange (ETDEWEB)

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

1969-03-01

A study has been made of the internal friction in different graphites between -180 C and +500 C using a torsion pendulum; the graphites had been previously treated thermo-mechanically, by neutron irradiation and subjected to partial annealings. It has been shown that there occurs: a hysteretic type dissipation of energy, connected with interactions between dislocations and other defects in the matrix; a dissipation having a partially hysteretic character which can be interpreted by a Granato-Luke type formalism and which is connected with the presence of an 'ultra-micro porosity'; a dissipation by a relaxation mechanism after a small dose of irradiation; this is attributed to the reorientation of bi-interstitials; a dissipation having the characteristics of a solid state transformation, this during an annealing after irradiation. It is attributed to the reorganization of interstitial defects. Some information has thus been obtained concerning graphites, in particular: their behaviour at low mechanical stresses, the nature of irradiation defects and their behaviour during annealing, the structural changes occurring during graphitization, the relationship between internal friction and macroscopic mechanical properties. (author) [French] L'etude du coefficient de frottement interieur au moyen d'un pendule de torsion entre -180 C et +500 C a ete realisee pour differents graphites apres des traitements thermo-mecaniques, des irradiations neutroniques et des guerisons partielles. Il a ete mis en evidence: une dissipation d'energie a caractere hysteretique, reliee aux interactions des dislocations avec les autres defauts de la matrice; une dissipation a caractere partiellement hysteretique, interpretable par un formalisme type Granato-Lucke et reliee a la presence d'une ''ultra-microporosite''; une dissipation par un mecanisme de relaxation, apres irradiation a faible dose, attribuee a la reorientation de di-interstitiels; une dissipation presentant les caracteristiques d

Characterisation of Chlorine Behavior in French Graphite

International Nuclear Information System (INIS)

Blondel, A.; Moncoffre, N.; Toulhoat, N.; Bererd, N.; Petit, L.; Laurent, G.; Lamouroux, C.

2016-01-01

Chlorine 36 is one of the main radionuclides of concern for French graphite waste disposal. In order to help the understanding of its leaching behaviour under disposal conditions, the respective impact of temperature, irradiation and gas radiolysis on chlorine release in reactor has been studied. Chlorine 36 has been simulated through chlorine 37 ion implantation in virgin nuclear graphite samples. Results show that part of chlorine is highly mobile in graphite in the range of French reactors operating temperatures in relation with graphite structural recovering. Ballistic damage generated by irradiation also promotes chlorine release whereas no clear impact of the coolant gas radiolysis was observed in the absence of graphite radiolytic corrosion. (author)

Growth of carbon nanotubes in arc plasma treated graphite disc: microstructural characterization and electrical conductivity study

Science.gov (United States)

Nayak, B. B.; Sahu, R. K.; Dash, T.; Pradhan, S.

2018-03-01

Circular graphite discs were treated in arc plasma by varying arcing time. Analysis of the plasma treated discs by field emission scanning electron microscope revealed globular grain morphologies on the surfaces, but when the same were observed at higher magnification and higher resolution under transmission electron microscope, growth of multiwall carbon nanotubes of around 2 nm diameter was clearly seen. In situ growth of carbon nanotube bundles/bunches consisting of around 0.7 nm tube diameter was marked in the case of 6 min treated disc surface. Both the untreated and the plasma treated graphite discs were characterized by X-ray diffraction, energy dispersive spectra of X-ray, X-ray photoelectron spectroscopy, transmission electron microscopy, micro Raman spectroscopy and BET surface area measurement. From Raman spectra, BET surface area and microstructure observed in transmission electron microscope, growth of several layers of graphene was identified. Four-point probe measurements for electrical resistivity/conductivity of the graphite discs treated under different plasma conditions showed significant increase in conductivity values over that of untreated graphite conductivity value and the best result, i.e., around eightfold increase in conductivity, was observed in the case of 6 min plasma treated sample exhibiting carbon nanotube bundles/bunches grown on disc surface. By comparing the microstructures of the untreated and plasma treated graphite discs, the electrical conductivity increase in graphite disc is attributed to carbon nanotubes (including bundles/bunches) growth on disc surface by plasma treatment.

Determination of Cl-36 in Irradiated Reactor Graphite

International Nuclear Information System (INIS)

Beer, H.-F.; Schumann, D.; Stowasser, T.; Hartmann, E.; Kramer, A.

2016-01-01

Two of the three research reactors at the Paul Scherrer Institute (PSI), the reactors DIORIT and PROTEUS, contained reactor graphite. Whereas the former research reactor DIORIT has been dismantled completely the PROTEUS is subject to a future decommissioning. In case of the DIORIT the reactor graphite was conditioned applying a procedure developed at PSI. In this case the 36 Cl content had to be determined after the conditioning. The result is reported in this paper. The radionuclide inventory including 36 Cl of the graphite used in PROTEUS was measured and the results are reported in here. It has been proven that the graphite from PROTEUS has a radionuclide inventory near the detection limits. All determined radionuclide activities are far below the Swiss exemptions limits. The graphite from PROTEUS therefore poses no radioactive waste. In contrast, the 36 Cl content of graphite from DIORIT is well above the exemption limits. (author)

Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors

International Nuclear Information System (INIS)

Vaudey, C.E.

2010-10-01

This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the French law of June 206. These wastes contain two long-lived radionuclides ( 14 C and 36 Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of 36 Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion independently. Our results show a rapid release of around 20% 36 Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic. (author)

Graphite targets at LAMPF

International Nuclear Information System (INIS)

Brown, R.D.; Grisham, D.L.

1983-01-01

Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

Comparison of 3 MeV C+ ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

International Nuclear Information System (INIS)

Chi, Se-Hwan; Kim, Gen-Chan

2008-01-01

Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ∼19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

Comparison of 3 MeV C + ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

Science.gov (United States)

Chi, Se-Hwan; Kim, Gen-Chan

2008-10-01

Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ˜19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

Synthesis of metal free ultrathin graphitic carbon nitride sheet for photocatalytic dye degradation of Rhodamine B under visible light irradiation

Science.gov (United States)

Rahman, Shakeelur; Momin, Bilal; Higgins M., W.; Annapure, Uday S.; Jha, Neetu

2018-04-01

In recent times, low cost and metal free photocatalyts driven under visible light have attracted a lot of interest. One such photo catalyst researched extensively is bulk graphitic carbon nitride sheets. But the low surface area and weak mobility of photo generated electrons limits its photocatalytic performance in the visible light spectrum. Here we present the facile synthesis of ultrathin graphitic carbon nitride using a cost effective melamine precursor and its application in highly efficient photocatalytic dye degradation of Rhodamine B molecules. Compared to bulk graphitic carbon nitride, the synthesized ultrathin graphitic carbon nitride shows an increase in surface area, a a decrease in optical band gap and effective photogenerated charge separation which facilitates the harvest of visible light irradiation. Due to these optimal properties of ultrathin graphitic carbon nitride, it shows excellent photocatalytic activity with photocatalytic degradation of about 95% rhodamine B molecules in 1 hour.

Impermeable Graphite: A New Development for Embedding Radioactive Waste

International Nuclear Information System (INIS)

Fachinger, Johannes

2016-01-01

Irradiated graphite has to be handled as radioactive waste after the operational period of the reactor. However, the waste management of irradiated graphite e.g. from the Spanish Vandellos reactor shows, that waste management of even low contaminated graphite could be expensive and requires special retrieval, treatment and disposal technologies for safe long term storage as low or medium radioactive waste. FNAG has developed an impermeable graphite matrix (IGM) as nuclear waste embedding material. This IGM provides a long term stable enclosure of radioactive waste and can reuse irradiated graphite as feedstock material. Therefore, no additional disposal volume is required if e.g. concrete waste packages were replaced by IGM waste packages. The variability of IGM as embedding has been summarized in the following paper usable for metal scraps, ion exchange resins or debris from buildings. Furthermore the main physical, chemical and structural properties are described. (author)

Accelerated irradiation growth of zirconium alloys

International Nuclear Information System (INIS)

Griffiths, M.; Gilbert, R.W.; Fidleris, V.

1989-01-01

This paper discusses how sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 10 25 n · m -2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or breakaway growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature. Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences ( 2 :5 n · m -2 , E > 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates. (author)

Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

Energy Technology Data Exchange (ETDEWEB)

Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

2002-12-01

The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

Synthesis of carbon-13 labelled carbonaceous deposits and their evaluation for potential use as surrogates to better understand the behaviour of the carbon-14-containing deposit present in irradiated PGA graphite

Energy Technology Data Exchange (ETDEWEB)

Payne, L., E-mail: liam.payne@bristol.ac.uk [Interface Analysis Centre, HH Wills Physics Laboratory, University of Bristol, BS8 1TL (United Kingdom); Walker, S.; Bond, G. [Centre for Materials Science, University of Central Lancashire, PR1 2HE (United Kingdom); Eccles, H. [John Tyndall Institute for Nuclear Research, School of Computing, Engineering and Physical Sciences, University of Central Lancashire, PR1 2HE (United Kingdom); Heard, P.J.; Scott, T.B. [Interface Analysis Centre, HH Wills Physics Laboratory, University of Bristol, BS8 1TL (United Kingdom); Williams, S.J. [Radioactive Waste Management, B587, Curie Avenue, Harwell Oxford, Didcot, OX11 0RH (United Kingdom)

2016-03-15

The present work has used microwave plasma chemical vapour deposition to generate suitable isotopically labelled carbonaceous deposits on the surface of Pile Grade A graphite for use as surrogates for studying the behaviour of the deposits observed on irradiated graphite extracted from UK Magnox reactors. These deposits have been shown elsewhere to contain an enhanced concentration of {sup 14}C compared to the bulk graphite. A combination of Raman spectroscopy, ion beam milling with scanning electron microscopy and secondary ion mass spectrometry were used to determine topography and internal morphology in the formed deposits. Direct comparison was made against deposits found on irradiated graphite samples trepanned from a Magnox reactor core and showed a good similarity in appearance. This work suggests that the microwave plasma chemical vapour deposition technique is of value in producing simulant carbon deposits, being of sufficiently representative morphology for use in non-radioactive surrogate studies of post-disposal behaviour of {sup 14}C-containing deposits on some irradiated Magnox reactor graphite.

Damaging process of graphite - new model and its impact on degradation of materials performance

International Nuclear Information System (INIS)

Tanabe, T.; Muto, S.

1999-01-01

The most widely accepted model for development of defect structure in neutron irradiated graphite has been such that following the first production of a pair of an interstitial and vacancy, di-interstitials and vacancies are formed and their subsequent growth would result in the production of an interstitial plane or loop in-between the basal planes and vacancy clusters, respectively, which could cause the loss of thermal conductivity and dimensional change. Recently we have claimed that the formation of vacancy clusters and growth of the interstitial planes are not necessarily a unique interpretation of the damaging process. Instead, the damaging process is described by orientational disordering within the basal planes, i.e. fragmentation into small crystallites and rotation of their crystalline axes, change of stacking order and elongation of the interplanar spacing. The orientational disordering within the basal planes proceeds coordinately over a few layers with their layered correlation maintained. This process accompanies changes in bonding nature producing 5 member- and 7 member-atomic rings as appeared in fullerenes. This is so to speak ''self-restoring or reconstruction'' to maintain resonance bonds as strict as possible without the formation of dangling bonds. This paper reviews irradiation effects in graphite such as increase of hydrogen retention, loss of thermal conductivity and dimensional change on the bases of our new model, taking account of the changes of the bonding nature in irradiated graphite. (orig.)

Status of IAEA international data base on irradiated graphite properties with respect to HTR engineering issues

International Nuclear Information System (INIS)

Hacker, P.J.; Haag, G.

2002-01-01

The International Database on Irradiated Nuclear Graphite Properties contains data on the physical, chemical, mechanical and other relevant properties of graphites. Its purpose is to provide a platform that makes these properties accessible to approved users in the fields of nuclear power, nuclear safety and other nuclear science and technology applications. The database is constructed using Microsoft Access 97 software and has a controlled distribution by CD ROM to approved users. This paper describes the organisation and management of the database through administrative arrangements approved by the IAEA. It also outlines the operation of the database. The paper concludes with some remarks upon and illustrations of the usefulness of the database for the design and operation of HTR. (authors)

Cyclic fatigue of near-isotopic graphite: influence of stress cycle and neutron irradiation

International Nuclear Information System (INIS)

Price, R.J.

1977-11-01

Near-isotropic graphites H-451 and PGX were tested in uniaxial cyclic fatigue, and fatigue life (S-N) curves were generated to a maximum of 10 5 cycles. The stress ratio, R (minimum stress during a cycle divided by maximum stress) ranged from -1 to +0.5. With R = - 1, the homologous stress limits (maximum applied fatigue stress divided by the tensile strength) for 50% specimen survival to 10 5 cycles averaged 0.63 in the axial direction and 0.74 in the radial direction. Corresponding homologous stress limits for 99% specimen survival (99/95 tolerance limits) were 0.48 and 0.53. Higher R-values resulted in longer fatigue lives and increased stress limits. H-451 graphite specimens irradiated with fast neutrons at 1173 to 1263 0 K at fluences of up to 10 26 n/m 2 (equivalent fission fluence) showed fatigue stress limits of about twice the unirradiated levels when the unirradiated tensile strength was used as the basis for normalization

HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

Energy Technology Data Exchange (ETDEWEB)

1977-09-01

The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

Graphite materials testing in the ATR for lifetime management of Magnox reactors

International Nuclear Information System (INIS)

Grover, S.B.; Metcalfe, M.P.

2002-01-01

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment. (author)

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors

International Nuclear Information System (INIS)

Grover, S.B.; Metcalfe, M.P.

2002-01-01

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment

Physics experiments in graphite lattices (1962); Experiences sur les reseaux a graphite (1962)

Energy Technology Data Exchange (ETDEWEB)

Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-07-01

A review is made of the various experimental methods used to determine the physics of graphite, natural uranium lattices: integral lattice experiments; both absolute and differential, effective cross section measurements, both by activation methods and by analysis of irradiated fuels, fine structure measurements. A number of experimental results are also given. (authors) [French] On decrit les differentes methodes experimentales utilisees pour determiner les parametres physiques de reseaux a uranium-graphite. Il s'agit d'experiences globales: mesures absolues et relatives de laplaciens, mesures de sections efficaces effectives par activation et par analyses de combustibles irradies, mesures de structures fines. Un certain nombre de resultats experimentaux sont communiques. (auteurs)

A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

International Nuclear Information System (INIS)

Gesh, Christopher J.

2004-01-01

The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

Contribution to the study of internal friction in graphites

International Nuclear Information System (INIS)

Merlin, J.

1969-03-01

A study has been made of the internal friction in different graphites between -180 C and +500 C using a torsion pendulum; the graphites had been previously treated thermo-mechanically, by neutron irradiation and subjected to partial annealings. It has been shown that there occurs: a hysteretic type dissipation of energy, connected with interactions between dislocations and other defects in the matrix; a dissipation having a partially hysteretic character which can be interpreted by a Granato-Luke type formalism and which is connected with the presence of an 'ultra-micro porosity'; a dissipation by a relaxation mechanism after a small dose of irradiation; this is attributed to the reorientation of bi-interstitials; a dissipation having the characteristics of a solid state transformation, this during an annealing after irradiation. It is attributed to the reorganization of interstitial defects. Some information has thus been obtained concerning graphites, in particular: their behaviour at low mechanical stresses, the nature of irradiation defects and their behaviour during annealing, the structural changes occurring during graphitization, the relationship between internal friction and macroscopic mechanical properties. (author) [fr

low dose irradiation growth in zirconium

International Nuclear Information System (INIS)

Fortis, A.M.

1987-01-01

Low dose neutron irradiation growth in textured and recrystallized zirconium, is studied, at the Candu Reactors Calandria temperature (340 K) and at 77 K. It was necessary to design and build 1: A facility to irradiate at high temperatures, which was installed in the Argentine Atomic Energy Commission's RA1 Reactor; 2: Devices to carry out thermal recoveries, and 3: Devices for 'in situ' measurements of dimensional changes. The first growth kinetics curves were obtained at 365 K and at 77 K in a cryostat under neutron fluxes of similar spectra. Irradiation growth experiments were made in zirconium doped with fissionable material (0,1 at % 235 U). In this way an equivalent dose two orders of magnitude greater than the reactor's fast neutrons dose was obtained, significantly reducing the irradiation time. The specimens used were bimetallic couples, thus obtaining a great accuracy in the measurements. The results allow to determine that the dislocation loops are the main cause of irradiation growth in recrystallized zirconium. Furthermore, it is shown the importance of 'in situ' measurements as a way to avoid the effect that temperature changes have in the final growth measurement; since they can modify the residual stresses and the overconcentrations of defects. (M.E.L.) [es

Sintered tantalum carbide coatings on graphite substrates: Highly reliable protective coatings for bulk and epitaxial growth

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Daisuke; Suzumura, Akitoshi; Shigetoh, Keisuke [Toyota Central R and D Labs., Inc., Nagakute, Aichi 480-1192 (Japan)

2015-02-23

Highly reliable low-cost protective coatings have been sought after for use in crucibles and susceptors for bulk and epitaxial film growth processes involving wide bandgap materials. Here, we propose a production technique for ultra-thick (50–200 μmt) tantalum carbide (TaC) protective coatings on graphite substrates, which consists of TaC slurry application and subsequent sintering processes, i.e., a wet ceramic process. Structural analysis of the sintered TaC layers indicated that they have a dense granular structure containing coarse grain with sizes of 10–50 μm. Furthermore, no cracks or pinholes penetrated through the layers, i.e., the TaC layers are highly reliable protective coatings. The analysis also indicated that no plastic deformation occurred during the production process, and the non-textured crystalline orientation of the TaC layers is the origin of their high reliability and durability. The TaC-coated graphite crucibles were tested in an aluminum nitride (AlN) sublimation growth process, which involves extremely corrosive conditions, and demonstrated their practical reliability and durability in the AlN growth process as a TaC-coated graphite. The application of the TaC-coated graphite materials to crucibles and susceptors for use in bulk AlN single crystal growth, bulk silicon carbide (SiC) single crystal growth, chemical vapor deposition of epitaxial SiC films, and metal-organic vapor phase epitaxy of group-III nitrides will lead to further improvements in crystal quality and reduced processing costs.

Sintered tantalum carbide coatings on graphite substrates: Highly reliable protective coatings for bulk and epitaxial growth

International Nuclear Information System (INIS)

Nakamura, Daisuke; Suzumura, Akitoshi; Shigetoh, Keisuke

2015-01-01

Highly reliable low-cost protective coatings have been sought after for use in crucibles and susceptors for bulk and epitaxial film growth processes involving wide bandgap materials. Here, we propose a production technique for ultra-thick (50–200 μmt) tantalum carbide (TaC) protective coatings on graphite substrates, which consists of TaC slurry application and subsequent sintering processes, i.e., a wet ceramic process. Structural analysis of the sintered TaC layers indicated that they have a dense granular structure containing coarse grain with sizes of 10–50 μm. Furthermore, no cracks or pinholes penetrated through the layers, i.e., the TaC layers are highly reliable protective coatings. The analysis also indicated that no plastic deformation occurred during the production process, and the non-textured crystalline orientation of the TaC layers is the origin of their high reliability and durability. The TaC-coated graphite crucibles were tested in an aluminum nitride (AlN) sublimation growth process, which involves extremely corrosive conditions, and demonstrated their practical reliability and durability in the AlN growth process as a TaC-coated graphite. The application of the TaC-coated graphite materials to crucibles and susceptors for use in bulk AlN single crystal growth, bulk silicon carbide (SiC) single crystal growth, chemical vapor deposition of epitaxial SiC films, and metal-organic vapor phase epitaxy of group-III nitrides will lead to further improvements in crystal quality and reduced processing costs

Modification of PMMA/graphite nanocomposites through ion beam technique

Science.gov (United States)

Singhal, Prachi; Rattan, Sunita; Avasthi, Devesh Kumar; Tripathi, Ambuj

2013-08-01

Swift heavy ion (SHI) irradiation is a special technique for inducing physical and chemical modifications in bulk materials. In the present work, the SHI hs been used to prepare nanocomposites with homogeneously dispersed nanoparticles. The nanographite was synthesized from graphite using the intercalation-exfoliation method. PMMA Poly(methyl methacrylate)/graphite nanocomposites have been synthesized by in situ polymerization. The prepared PMMA/graphite nanocomposite films were irradiated with SHI irradiation (Ni ion beam, 80 MeV and C ion beam, 50 MeV) at a fluence of 1×1010 to 3×1012 ions/cm2. The nanocomposite films were characterized by scanning electron microscope (SEM) and were evaluated for their electrical and sensor properties. After irradiation, significant changes in surface morphology of nanocomposites were observed as evident from the SEM images, which show the presence of well-distributed nanographite platelets. The irradiated nanocomposites exhibit better electrical and sensor properties for the detection of nitroaromatics with marked improvement in sensitivity as compared with unirradiated nanocomposites.

{sup 36}Cl and {sup 14}C behaviour in UNGG graphite during leaching experiments

Energy Technology Data Exchange (ETDEWEB)

Pichon, C.; Guy, C.; Comte, J. [Commissariat a l' Energie Atomique - C.E.A., Laboratoire d' Analyses Radiochimiques et Chimiques (L.A.R.C.) 13108 Saint Paul lez Durance (France)

2008-07-01

Graphite has been used as a moderator in Natural Uranium Graphite Gas reactors. Among the radionuclides, the long-lived activation product {sup 36}Cl and {sup 14}C, which are abundant in graphite after irradiation can be the main contributors to the dose during disposal. This paper deals with the first results obtained on irradiated graphite from French G2 reactor. Both leaching and diffusion experiments have been performed in order to understand and quantify the radionuclides behaviour. Chlorine leaching seems to be controlled by diffusion transport through graphite matrix. On the contrary {sup 14}C leaching is very low, probably because after irradiation, the remaining {sup 14}C was produced from {sup 13}C activation in the crystalline structure of graphite. (authors)

Time of flight measurements of unirradiated and irradiated nuclear graphite under cyclic compressive load

Energy Technology Data Exchange (ETDEWEB)

Bodel, W., E-mail: william.bodel@hotmail.com [Nuclear Graphite Research Group, The University of Manchester (United Kingdom); Atkin, C. [Health and Safety Laboratory, Buxton (United Kingdom); Marsden, B.J. [Nuclear Graphite Research Group, The University of Manchester (United Kingdom)

2017-04-15

The time-of-flight technique has been used to investigate the stiffness of nuclear graphite with respect to the grade and grain direction. A loading rig was developed to collect time-of-flight measurements during cycled compressive loading up to 80% of the material's compressive strength and subsequent unloading of specimens along the axis of the applied stress. The transmission velocity (related to Young's modulus), decreased with increasing applied stress; and depending on the graphite grade and orientation, the modulus then increased, decreased or remained constant upon unloading. These tests were repeated while observing the microstructure during the load/unload cycles. Initial decreases in transmission velocity with compressive load are attributed to microcrack formation within filler and binder phases. Three distinct types of behaviour occur on unloading, depending on the grade, irradiation, and loading direction. These different behaviours can be explained in terms of the material microstructure observed from the microscopy performed during loading.

Characteristics of first loaded IG-110 graphite in HTTR core

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

2006-10-01

IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

Diamond growth on Fe-Cr-Al alloy by H2-plasma enhanced graphite etching

International Nuclear Information System (INIS)

Li, Y. S.; Hirose, A.

2007-01-01

Without intermediate layer and surface pretreatment, adherent diamond films with high initial nucleation density have been deposited on Fe-15Cr-5Al (wt. %) alloy substrate. The deposition was performed using microwave hydrogen plasma enhanced graphite etching in a wide temperature range from 370 to 740 degree sign C. The high nucleation density and growth rate of diamond are primarily attributed to the unique precursors used (hydrogen plasma etched graphite) and the chemical nature of the substrate. The improvement in diamond adhesion to steel alloys is ascribed to the important role played by Al, mitigation of the catalytic function of iron by suppressing the preferential formation of loose graphite intermediate phase on steel surface

Voronoi-Tessellated Graphite Produced by Low-Temperature Catalytic Graphitization from Renewable Resources.

Science.gov (United States)

Zhao, Leyi; Zhao, Xiuyun; Burke, Luke T; Bennett, J Craig; Dunlap, Richard A; Obrovac, Mark N

2017-09-11

A highly crystalline graphite powder was prepared from the low temperature (800-1000 °C) graphitization of renewable hard carbon precursors using a magnesium catalyst. The resulting graphite particles are composed of Voronoi-tessellated regions comprising irregular sheets; each Voronoi-tessellated region having a small "seed" particle located near their centroid on the surface. This suggests nucleated outward growth of graphitic carbon, which has not been previously observed. Each seed particle consists of a spheroidal graphite shell on the inside of which hexagonal graphite platelets are perpendicularly affixed. This results in a unique high surface area graphite with a high degree of graphitization that is made with renewable feedstocks at temperatures far below that conventionally used for artificial graphites. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Management of UKAEA graphite liabilities

International Nuclear Information System (INIS)

Wise, M.

2001-01-01

The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the

Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

Directory of Open Access Journals (Sweden)

Bespala Evgeny

2016-01-01

Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.

Graphite behaviour in relation to the fuel element design

Energy Technology Data Exchange (ETDEWEB)

Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

1969-09-01

The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 10²¹ cm ^-2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.

Experience with graphite in JET

International Nuclear Information System (INIS)

Pick, M.A.; Celentano, G.; Deksnis, E.; Dietz, K.J.; Shaw, R.; Sonnenberg, K.; Walravens, M.

1987-01-01

During the current operational period of JET more than 50% of the internal area of the machine is covered in graphite tiles. This includes the 15 m 2 of carbon tiles installed in the new toroidal limiter, the 40 poloidal belts of graphite tiles covering the U-joints and bellows as well as a two metre high ring (-- 20 m 2 ) or carbon tiles on the inner wall of the Torus. A ring of tiles in the equatorial plane (3 tiles high) consists of carbon-carbon fibre tiles. Test bed results indicated that the fine grained graphite tiles cracked at ∼ 1 kW/cm 2 for 2s of irradiation whereas the carbon-carbon fibre tiles were able to sustain a flux, limited by the irradiation facility, of 3.5 kW for 3s without any damage. The authors report on the generally positive experience they have had had with the installed graphite during the present and previous in-vessel configurations. This includes the physical integrity of the tiles under severe conditions such as high energy run-away electron beams, plasma disruptions and high heat fluxes. They report on the importance of the precise positioning of the inner wall and x-point tiles at the very high power fluxes of JET and the effect of deviations on both graphite and carbon-fibre tiles

Chemical atomization of graphite by H+ ions

International Nuclear Information System (INIS)

Busharov, I.P.; Gorbatov, E.A.; Gusev, V.M.; Guseva, M.I.; Martynenko, Yu.V.

A simple model of the mechanism of chemical atomization is given, on whose basis a decrease in chemical atomization is qualitatively predicted for high temperatures. Mass spectrometric investigations of the atomization products cited, which found CH 4 and CH 3 molecules during the irradiation of graphite and H + ions thereby confirmed the presence of chemical atomization. A relationship of S and temperature of graphite T during irradiation was obtained which showed a decrease in the coefficient of atomization of a high temperature. (U.S.)

Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-08-01

The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs.

Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

International Nuclear Information System (INIS)

1996-08-01

The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs

Antibacterial activity of irradiated and non-irradiated chitosan and chitosan derivatives against Escherichia coli growth

International Nuclear Information System (INIS)

Tg Ahbrizal Farizal Tg Ahmad; Norimah Yusof; Kamarudin Bahari; Kamaruddin Hashim

2006-01-01

Samples of chitosan and four chitosan derivatives [ionic chitosan, chitosan lactate, carboxymethyl chitosan (C) and carboxymethyl chitosan (L)] were studied for their antibacterial activities against Escherichia coli growth. Chitosan and chitosan derivatives were prepared at concentrations 20, 100, 1000, 10000 ppm and 250, 1000, 5000, 10000, 20000 ppm, respectively. Each of the samples was tested before and after irradiation with electron beam at 25 kGy. The turbidity of bacterial growth media was measured periodically at 0, 0.5, 1, 2, 4, 6 and 24 h after inoculation using the optical density method. The results indicated that non- irradiated chitosan inhibited E. coli growth at 20 and 100 ppm. Meanwhile, irradiated chitosan at 100 and 1000 ppm concentration inhibited E. coli growth. Both irradiated and non-irradiated ionic chitosan inhibited E. coli growth at all concentrations used. Chitosan lactate was found to inhibit E. coli at concentration as low as 5000 ppm for both irradiated and non-irradiated samples. E. coli growth was not inhibited by carboxymethyl chitosan (C) and carboxymethyl chitosan (L), before and after irradiation. The findings suggested that chitosan has greater antibacterial activity as compared to the chitosan derivative samples. (Author)

Methodology of characterization of radioactive graphite

International Nuclear Information System (INIS)

Pina, G.; Rodriguez, M.; Lara, E.; Magro, E.; Gascon, J. L.; Leganes, J. L.

2014-01-01

Since the dismantling of Vandellos I, ENRESA has promoted the precise knowledge of the inventory of irradiated graphite (graphite-i) through establishing methodologies for radiological characterization of the vector of radionuclides of interest and their correlations as the primary means of characterization strategy to establish the safer management of this material in its life cycle. (Author)

Heavy-ion irradiation induced diamond formation in carbonaceous materials

International Nuclear Information System (INIS)

Daulton, T. L.

1999-01-01

The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond

Assessment of management modes for graphite from reactor decommissioning

International Nuclear Information System (INIS)

White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

1984-01-01

A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

Surface structure modification of single crystal graphite after slow, highly charged ion irradiation

Science.gov (United States)

Alzaher, I.; Akcöltekin, S.; Ban-d'Etat, B.; Manil, B.; Dey, K. R.; Been, T.; Boduch, P.; Rothard, H.; Schleberger, M.; Lebius, H.

2018-04-01

Single crystal graphite was irradiated by slow, highly charged ions. The modification of the surface structure was studied by means of Low-Energy Electron Diffraction. The observed damage cross section increases with the potential energy, i.e. the charge state of the incident ion, at a constant kinetic energy. The potential energy is more efficient for the damage production than the kinetic energy by more than a factor of twenty. Comparison with earlier results hints to a strong link between early electron creation and later target atom rearrangement. With increasing ion fluence, the initially large-scale single crystal is first transformed into μ m-sized crystals, before complete amorphisation takes place.

The irradiation behaviour of boron carbide/graphite between 800 and 1,1000C

International Nuclear Information System (INIS)

Hattenbach, K.; Hilgendorff, W.; Weiler, K.; Zimmermann, H.U.

1975-01-01

64 samples of boron carbide/graphite, a material used as burnable poison in high temperature reactors, were irradiated at temperatures between 800 and 1,100 0 C up to a fluence of 1-2 x 10 20 nvt. The following post-investigations were extended to dimensional measurements to determime a possible swelling or shrinking of the pellet, corrosion tests in completely desalinated water at 300 0 C, preparation of metallographic microsections to check for crack formation, determination of the helium hold back power and the thus involved gas chromatic analysis, as well as burn-up determinations by determining the boron 10/boron 11 ratio and the lithium concentration. (orig./LN) [de

Graphite target for the spiral project

International Nuclear Information System (INIS)

Putaux, J.C.; Ducourtieux, M.; Ferro, A.; Foury, P.; Kotfila, L.; Mueller, A.C.; Obert, J.; Pauwels, N.; Potier, J.C.; Proust, J.; Loiselet, M.

1996-01-01

A study of the thermal and physical properties of graphite targets for the SPIRAL project is presented. The main objective is to develop an optimized set-up both mechanically and thermally resistant, presenting good release properties (hot targets with thin slices). The results of irradiation tests concerning the mechanical and thermal resistance of the first prototype of SPIRAL target with conical geometry are presented. The micro-structural properties of the graphite target is also studied, in order to check that the release properties are not deteriorated by the irradiation. Finally, the results concerning the latest pilot target internally heated by an electrical current are shown. (author)

Surface analysis of graphite fiber reinforced polyimide composites

Science.gov (United States)

Messick, D. L.; Progar, D. J.; Wightman, J. P.

1983-01-01

Several techniques have been used to establish the effect of different surface pretreatments on graphite-polyimide composites. Composites were prepared from Celion 6000 graphite fibers and the polyimide LARC-160. Pretreatments included mechanical abrasion, chemical etching and light irradiation. Scanning electron microscopy (SEM) and X-ray photoelectron spectroscopy (XPS) were used in the analysis. Contact angle of five different liquids of varying surface tensions were measured on the composites. SEM results showed polymer-rich peaks and polymer-poor valleys conforming to the pattern of the release cloth used durng fabrication. Mechanically treated and light irradiated samples showed varying degrees of polymer peak removal, with some degradation down to the graphite fibers. Minimal changes in surface topography were observed on concentrations of surface fluorine even after pretreatment. The light irradiation pretreatment was most effective at reducing surface fluorine concentrations whereas chemical pretreatment was the least effective. Critical surface tensions correlated directly with the surface fluorine to carbon ratios as calculated from XPS.

On the defect structure due to low energy ion bombardment of graphite

Science.gov (United States)

Marton, D.; Bu, H.; Boyd, K. J.; Todorov, S. S.; Al-Bayati, A. H.; Rabalais, J. W.

1995-03-01

Graphite surfaces cleaved perpendicular to the c axis have been irradiated with low doses of Ar + ions at 50 eV kinetic energy and perpendicular incidence. Scanning tunneling micrographs (STM) of these irradiated surfaces exhibited dome-like features as well as point defects. These dome-like features retain undisturbed graphite periodicity. This finding is attributed to the stopping of ions between the first and second graphite sheets. The possibility of doping semiconductors at extremely shallow depths is raised.

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

International Nuclear Information System (INIS)

Fachinger, Johannes; Muller, Walter; Marsat, Eric; Grosse, Karl-Heinz; Seemann, Richard; Scales, Charlie; Easton, Michael Mark; Anthony Banford

2013-01-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

Energy Technology Data Exchange (ETDEWEB)

Fachinger, Johannes; Muller, Walter [FNAG ZU Hanau, Hanau (Germany); Marsat, Eric [FNAG SAS Le Pont de Claix (France); Grosse, Karl-Heinz; Seemann, Richard [ALD Hanau (Germany); Scales, Charlie; Easton, Michael Mark [NNL, Workington (United Kingdom); Anthony Banford [NNL, Warrington (United Kingdom); University of Manchester, Manchester (United Kingdom)

2013-07-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

In-situ electric resistance measurements and annealing effects of graphite exposed to swift heavy ions

International Nuclear Information System (INIS)

Fernandes, Sandrina; Pellemoine, Frederique; Tomut, Marilena; Avilov, Mikhail; Bender, Markus; Boulesteix, Marine; Krause, Markus; Mittig, Wolfgang; Schein, Mike; Severin, Daniel; Trautmann, Christina

2013-01-01

To study the suitability of using graphite as material for high-power targets for rare isotope production at the future Facility for Rare Isotope Beams (FRIB) in the USA and at the Facility for Antiproton and Ion Research (FAIR) in Germany, thin foils of polycrystalline graphite were exposed to 8.6-MeV/u Au ions reaching a maximum fluence of 1 × 10 15 ions/cm 2 . Foil irradiation temperatures of up to 1800 °C were obtained by ohmic heating. In-situ monitoring of the electrical resistance of the graphite foils during and after irradiation provided information on beam-induced radiation damage. The rate of electrical resistance increase as a function of fluence was found to decrease with increasing irradiation temperature, indicating a more efficient annealing of the irradiation-produced defects. This is corroborated by the observation that graphite foils irradiated at temperatures below about 800 °C showed cracks and pronounced deformations, which did not appear on the samples irradiated at higher temperatures

Strategy for Handling and Treatment of INPP RBMK-1500 Irradiated Graphite

International Nuclear Information System (INIS)

Oryšaka, A.

2016-01-01

There are two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at Ignalina NPP. After the final shutdown of the INPP, radioactive i-graphite dismantling, handling, conditioning, storage and disposal is an important part of the decommissioning activities. The core of the INPP unit 1 and 2 contains about 3600 tons of i-graphite. Formation of activation products strongly depends on the contents of impurities, operational mode and concentration of impurities in the graphite. The case study for INPP envisages the analysis of possibilities of graphite handling and treatment in the context of immediate decommissioning. (author)

Effect of graphite reflector on activation of fusion breeding blanket

International Nuclear Information System (INIS)

Lee, Cheol Woo; Lee, Young-Ouk; Lee, Dong Won; Cho, Seungyon; Ahn, Mu-Young

2016-01-01

Highlights: • The graphite reflector concept has been applied in the design of the Korea HCCR TBM for ITER and this concept is also a candidate design option for Korea Demo. • In the graphite reflector, C-14, B-11 and Be-10 are produced after an irradiation. Impurities in both case of beryllium and graphite is dominant in the shutdown dose after an irradiation. • Based on the evaluation, the graphite reflector is a good alternative of the beryllium multiplier in the view of induced activity and shutdown dose. But C-14 produced in the graphite reflector should be considered carefully in the view of radwaste management. - Abstract: Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. In this paper, activity analysis was performed and the effect of graphite reflector in the view of activation was compared to the beryllium multiplier. As a result, it is expected that using the graphite reflector instead of the beryllium multiplier decreases total activity very effectively. But the graphite reflector produces C-14 about 17.2 times than the beryllium multiplier. Therefore, C-14 produced in the graphite reflector is expected as a significant nuclide in the view of radwaste management.

Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming

International Nuclear Information System (INIS)

Silbermann, Gwennaelle

2013-01-01

The dismantling of UNGG reactors in France will generate about 23 000 tons of radioactive graphite wastes. To manage these wastes, the radiological inventory and data on radionuclides (RN) location and speciation should be determined. 14 C was identified as an important RN for disposal due to its high initial activity and the risk of release of a mobile organic fraction in environment, after water ingress into the disposal. Hence, the objective of this thesis, carried out in partnership with EDF is to implement experimental studies to simulate and evaluate the impact of temperature, irradiation and graphite radiolytic corrosion on the in reactor behavior of 14 C and its precursor, 14 N. The obtained data are then used to study the thermal decontamination of graphite in presence of water vapor. The experimental approach aims at simulating the presence of 14 C and 14 N by the respective ion implantation of 13 C and 14 N or 15 N in virgin graphite. This study shows that, in the temperature range reached during reactor operation, (100-500 C) and without radiolytic corrosion, 13 C is thermally stable whatever the initial graphite structure. Moreover, irradiation experiments were performed on heated graphite (500 C) put in contact with a gas representative of the radiolized coolant gas. They show the synergistic role played by the oxidative species and the graphite structure disorder on the enhancement of 13 C mobility resulting in the gasification of the graphite surface and/or the selective oxidation of 13 C more weakly bound than 12 C. Concerning the pristine nitrogen, we showed first that the surface concentration reaches several hundred ppm (≤500 ppm at) and decreases at deeper depths to about 160 ppm at.. Unlike implanted 13 C, implanted nitrogen migrates at 500 C when the graphite is highly disordered (about 8 dpa) while remaining stable for a lower disorder rate (0.14 dpa). Experiments also show the synergistic role by electronic excitations and temperature

Models of bending strength for Gilsocarbon graphites irradiated in inert and oxidising environments

International Nuclear Information System (INIS)

Eason, Ernest D.; Hall, Graham N.; Marsden, Barry J.; Heys, Graham B.

2013-01-01

This paper presents the development and validation of an empirical model of fast neutron damage and radiolytic oxidation effects on bending strength for the moulded Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The inert environment model is based on evidence of essentially constant strength as fast neutron dose increases in inert environment. The model of combined irradiation and oxidation calibrates that constant along with an exponential function representing the degree of radiolytic oxidation as measured by weight loss. The change in strength with exposure was found to vary from one AGR station to another. The model was calibrated to data on material trepanned from AGR moderator bricks after varying operating times

Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

2016-04-15

Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.

Modification of cell growth rate by irradiation

International Nuclear Information System (INIS)

Itoh, Hisao; Takemasa, Kazuhiko; Nishiguchi, Iku; Ka, Wei-Jei; Kutsuki, Shoji; Hashimoto, Shozo

1993-01-01

The effect of irradiation on the proliferation kinetics of the monolayer cells has been studied. Two human cell lines with different doubling times (HeLa-P and RMUG) and two clones that have the same radiosensitivity but different doubling times (HeLa-R and HeLa-S) were irradiated with a daily dose of 2 Gy for 6 days. The number of the clonogenic cells/dish was calculated by multiplying the number of total cell/dish by the survival fraction. In the rapidly growing cells (HeLa-P, HeLa-R), the number of the clonogenic cells was not decreased by the first two fractionated irradiations, but decreased thereafter at a similar rate as by single-dose fractionation, whereas the clonogenic cell number decreased from the first fractionated irradiation in the slowly growing cells (RMUG, HeLa-S). When the proliferation of clonogenic cell number increased along with a similar growth rates that was seen in all other types of cells. Further, no correlation was seen between the growth rates of cells without irradiation and cells that received irradiation. This latter result suggests that the slow growth rate of non-irradiated cells may not be the predictive factor of the tumor cure and the interruption of radiotherapy may reduce the beneficial effect of this treatment even in slow growing tumors. (author)

In-situ electric resistance measurements and annealing effects of graphite exposed to swift heavy ions

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Sandrina [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Pellemoine, Frederique, E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Tomut, Marilena [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); National Institute for Materials Physics (NIMP), Bucharest (Romania); Avilov, Mikhail [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Bender, Markus [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Boulesteix, Marine [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Krause, Markus [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Technische Universität, Darmstadt (Germany); Mittig, Wolfgang [National Superconducting Cyclotron Lab (NSCL), Michigan State University, East Lansing, MI (United States); Schein, Mike [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Severin, Daniel [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Trautmann, Christina [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Technische Universität, Darmstadt (Germany)

2013-11-01

To study the suitability of using graphite as material for high-power targets for rare isotope production at the future Facility for Rare Isotope Beams (FRIB) in the USA and at the Facility for Antiproton and Ion Research (FAIR) in Germany, thin foils of polycrystalline graphite were exposed to 8.6-MeV/u Au ions reaching a maximum fluence of 1 × 10{sup 15} ions/cm{sup 2}. Foil irradiation temperatures of up to 1800 °C were obtained by ohmic heating. In-situ monitoring of the electrical resistance of the graphite foils during and after irradiation provided information on beam-induced radiation damage. The rate of electrical resistance increase as a function of fluence was found to decrease with increasing irradiation temperature, indicating a more efficient annealing of the irradiation-produced defects. This is corroborated by the observation that graphite foils irradiated at temperatures below about 800 °C showed cracks and pronounced deformations, which did not appear on the samples irradiated at higher temperatures.

A German research project about applicable graphite cutting techniques

International Nuclear Information System (INIS)

Holland, D.; Quade, U.; Bach, F.W.; Wilk, P.

2001-01-01

In Germany, too, quite large quantities of irradiated nuclear graphite, used in research and prototype reactors, are waiting for an environmental way of disposal. While incineration of nuclear graphite does not seem to be a publicly acceptable way, cutting and packaging into ductile cast iron containers could be a suitable way of disposal in Germany. Nevertheless, the cutting of graphite is also a very difficult technique by which a large amount of secondary waste or dust might occur. An applicable graphite cutting technique is needed. Therefore, a group of 13 German partners, consisting of one university, six research reactor operators, one technical inspection authority, three engineering companies, one industrial cutting specialist and one commercial dismantling company, decided in 1999 to start a research project to develop an applicable technique for cutting irradiated nuclear graphite. Aim of the project is to find the most suitable cutting techniques for the existing shapes of graphite blocks with a minimum of waste production rate. At the same time it will be learned how to sample the dust and collect it in a filter system. The following techniques will be tested and evaluated: thermal cutting, water jet cutting, mechanical cutting with a saw, plasma arc cutting, drilling. The subsequent evaluation will concentrate on dust production, possible irradiation of staff, time and practicability under different constraints. This research project is funded by the German Minister of Education and Research under the number 02 S 7849 for a period of two years. A brief overview about the work to be carried out in the project will be given. (author)

Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

International Nuclear Information System (INIS)

Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

1984-05-01

Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

Summary report of consultants' meeting on IAEA international database on irradiated nuclear graphite properties

International Nuclear Information System (INIS)

Humbert, D.; Wickham, A.J.

2007-06-01

The '9th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 26-27 March 2007 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

Role of growth hormone in stunted head growth after cranial irradiation

Energy Technology Data Exchange (ETDEWEB)

Clayton, P E; Shalet, S M; Price, D A; Surtees, R A; Pearson, D

1987-10-01

The head sizes of 38 patients, growth hormone (GH) deficient following craniospinal (n = 26) or cranial irradiation (n = 12), have been assessed before (n = 38) and on completion of GH therapy (n = 15) or at the end of a similar period of observation without GH (n = 7). These results were compared to the change in head size seen in idiopathic GH deficiency following GH therapy (n = 14). Before GH therapy, the latter had small heads (mean occipitofrontal circumference SD score (SDS) -1), which were relatively large compared to the height deficit (height SDS (CA) -4.7), and they exhibited catch-up growth with GH (delta occipitofrontal circumference SDS + 0.7, final occipitofrontal circumference SDS -0.2). In contrast, over a similar period all patients, who previously had received cranial irradiation in the dosage range 2700-4750 centi-Geigy, irrespective of the radiation schedule or GH treatment, showed a decrease in occipitofrontal circumference SDS (mean delta -0.9), a significant difference to the expected head growth of normal children over a similar period (p less than 0.01). We have noted that restricted head growth occurs in the years following cranial irradiation and is unaffected by GH therapy. Earlier work has shown that cranial irradiation may impair intelligence. The exact relationship between intellectual impairment and stunted head growth remains to be determined.

Modeling of irradiated graphite {sup 14}C transfer through engineered barriers of a generic geological repository in crystalline rocks

Energy Technology Data Exchange (ETDEWEB)

Poskas, Povilas; Grigaliuniene, Dalia, E-mail: Dalia.Grigaliuniene@lei.lt; Narkuniene, Asta; Kilda, Raimondas; Justinavicius, Darius

2016-11-01

There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of {sup 14}C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the {sup 14}C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released {sup 14}C into organic and inorganic compounds as well as the most recent information on {sup 14}C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic {sup 14}C into the geosphere can vary from 10{sup −} {sup 11} y{sup −} {sup 1} (for non-encapsulated graphite) to 10{sup −} {sup 12} y{sup −} {sup 1} (for encapsulated graphite) while of organic {sup 14}C it was about 10{sup −} {sup 3} y{sup −} {sup 1} of its inventory. Such difference demonstrates that investigations on the {sup 14}C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic {sup 14}C transfer was the sorption coefficient in the backfill and for organic {sup 14}C transfer – the backfill hydraulic conductivity. - Highlights: • Graphite moderated nuclear reactors are being decommissioned. • We studied interaction of disposed material with surrounding environment. • Specifically {sup 14}C transfer through engineered barriers of a geological repository. • Organic {sup 14}C flux to geosphere is considerably higher than inorganic

Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

International Nuclear Information System (INIS)

Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

2005-01-01

Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example

Micro-orientation control of silicon polymer thin films on graphite surfaces modified by heteroatom doping

Energy Technology Data Exchange (ETDEWEB)

Shimoyama, Iwao, E-mail: shimoyama.iwao@jaea.go.jp [Material Science Research Center, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan); Baba, Yuji [Fukushima Administrative Department, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan); Hirao, Norie [Material Science Research Center, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan)

2017-05-31

Highlights: • Micro-orientation control method for organic polysilane thin films is proposed. • This method utilizes surface modification of graphite using heteroatom doping. • Lying, standing, and random orientations can be freely controlled by this method. • Micro-pattering of a polysilane film with controlled orientations is achieved. - Abstract: Near-edge X-ray absorption fine structure (NEXAFS) spectroscopy is applied to study orientation structures of polydimethylsilane (PDMS) films deposited on heteroatom-doped graphite substrates prepared by ion beam doping. The Si K-edge NEXAFS spectra of PDMS show opposite trends of polarization dependence for non irradiated and N{sub 2}{sup +}-irradiated substrates, and show no polarization dependence for an Ar{sup +}-irradiated substrate. Based on a theoretical interpretation of the NEXAFS spectra via first-principles calculations, we clarify that PDMS films have lying, standing, and random orientations on the non irradiated, N{sub 2}{sup +}-irradiated, and Ar{sup +}-irradiated substrates, respectively. Furthermore, photoemission electron microscopy indicates that the orientation of a PDMS film can be controlled with microstructures on the order of μm by separating irradiated and non irradiated areas on the graphite surface. These results suggest that surface modification of graphite using ion beam doping is useful for micro-orientation control of organic thin films.

Scanning ion irradiation of polyimide films

Energy Technology Data Exchange (ETDEWEB)

Luecken, Stefan; Koval, Yuri; Mueller, Paul [Department of Physics and Interdisciplinary Center for Molecular Materials (ICMM), Universitaet Erlangen-Nuernberg (Germany)

2012-07-01

Recently we found, that the surface of nearly any polymer can be converted into conductive material by low energy ion irradiation. The graphitized layer consists of nanometer sized graphene and graphite flakes. In order to enhance the conductivity and to increase the size of the flakes we applied a novel method of scanning irradiation. We investigated the influence of various irradiation parameters on the conductivity of the graphitized layer. We show, that the conductance vs. temperature can be described in terms of weak Anderson localization. At approximately 70 K, a crossover occurs from 2-dimensional to 3-dimensional behavior. This can be explained by a decrease of the Thouless length with increasing temperature. The crossover temperature can be used to estimate the thickness of the graphitized layer.

Mode II interlaminar fracture of graphite/epoxy and graphite/PEEK

Science.gov (United States)

Carlsson, L. A.; Gillespie, J. W.; Trethewey, B. R.

1986-01-01

The end notched flexure (ENF) specimen is employed in an investigation of the interlaminar fracture toughness in Mode II (skew symmetric shear) loading of unidirectional graphite/epoxy and graphite/PEEK composites. Important experimental parameters such as the influence of precracking and the data reduction scheme for the Mode II toughness are discussed. Nonlinear load-deflection response is significant for the tough thermoplastic resin composite but is also present for the brittle thermoset composite. The observed nonlinearities, which are highly rate dependent, are attributed to a combination of slow stable crack growth preceding unstable crack growth and material inelastic behavior in the process zone around the crack tip.

Crystal plasticity modeling of irradiation growth in Zircaloy-2

Science.gov (United States)

Patra, Anirban; Tomé, Carlos N.; Golubov, Stanislav I.

2017-08-01

A physically based reaction-diffusion model is implemented in the visco-plastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. The reaction-diffusion model accounts for the defects produced by the cascade of displaced atoms, their diffusion to lattice sinks and the contribution to crystallographic strain at the level of single crystals. The VPSC framework accounts for intergranular interactions and irradiation creep, and calculates the strain in the polycrystalline ensemble. A novel scheme is proposed to model the simultaneous evolution of both, number density and radius, of irradiation-induced dislocation loops directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behaviour of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture and external stress on the coupled irradiation growth and creep behaviour are also studied and compared with available experimental data.

Fabrication of irradiation capsule for IASCC irradiation tests (1). Irradiation capsule for crack growth test (Joint research)

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Ishitsuka, Etsuo; Kawamura, Hiroshi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi

2008-03-01

It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, it is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported. (author)

Effect of High Energy Radiation on Mechanical Properties of Graphite Fiber Reinforced Composites. M.S. Thesis

Science.gov (United States)

Naranong, N.

1980-01-01

The flexural strength and average modulus of graphite fiber reinforced composites were tested before and after exposure to 0.5 Mev electron radiation and 1.33 Mev gamma radiation by using a three point bending test (ASTM D-790). The irradiation was conducted on vacuum treated samples. Graphite fiber/epoxy (T300/5208), graphite fiber/polyimide (C6000/PMR 15) and graphite fiber/polysulfone (C6000/P1700) composites after being irradiated with 0.5 Mev electron radiation in vacuum up to 5000 Mrad, show increases in stress and modulus of approximately 12% compared with the controls. Graphite fiber/epoxy (T300/5208 and AS/3501-6), after being irradiated with 1.33 Mev gamma radiation up to 360 Mrads, show increases in stress and modulus of approximately 6% at 167 Mrad compared with the controls. Results suggest that the graphite fiber composites studied should withstand the high energy radiation in a space environment for a considerable time, e.g., over 30 years.

Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

Energy Technology Data Exchange (ETDEWEB)

Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

2015-06-15

In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

Irradiation creep and growth of zircaloy-4 tubes

International Nuclear Information System (INIS)

Lansiart, S.; Darchis, L.; Pelchat, J.

1990-01-01

The influence of temperature and fast neutron flux on irradiation creep and growth of stress relieved zircaloy-4 pressurized tubes has been derived from experimental irradiations in NaK, performed up to 2.5 10 25 n.m -2 in the temperature range [280, 350] 0 C. A significant influence of temperature on axial growth has been observed: at 280 0 C the elongation can no longer be expressed as a linear function of fluence as for the 350 0 C irradiation temperature; diametral growth, on the other hand, always appears negligible. Irradiation creep obviously depends on temperature too; the diametral strain (including thermal part) has been modelled as a sum of primary and secondary terms, the former being independent of fluence. For the tubing considered it is observed that the ranking of the different batches, with respect to diametral creep resistance, is the same before and under irradiation. Concerning axial creep strain the stress relieved material behaves as does an isotropic tube. This is not the case of recrystallized zircaloy-4 F, which shows a non negligible axial deformation, related to the diametral creep one, even though this diametral irradiation creep strain is strongly reduced comparatively to that of the stress relieved material. The comparison of the two materials growth rates is more complex since their dependence on temperature and flux differs

Graphite surveillance in N Reactor

International Nuclear Information System (INIS)

Woodruff, E.M.

1991-09-01

Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

Graphite crystals grown within electromagnetically levitated metallic droplets

International Nuclear Information System (INIS)

Amini, Shaahin; Kalaantari, Haamun; Mojgani, Sasan; Abbaschian, Reza

2012-01-01

Various graphite morphologies were observed to grow within the electromagnetically levitated nickel–carbon melts, including primary flakes and spheres, curved surface graphite and eutectic flakes, as well as engulfed and entrapped particles. As the supersaturated metallic solutions were cooled within the electromagnetic (EM) levitation coil, the primary graphite flakes and spheres formed and accumulated near the periphery of the droplet due to EM circulation. The primary graphite islands, moreover, nucleated and grew on the droplet surface which eventually formed a macroscopic curved graphite crystal covering the entire liquid. Upon further cooling, the liquid surrounding the primary graphite went under a coupled eutectic reaction while the liquid in the center formed a divorced eutectic due to EM mixing. This brought about the formation of graphite fine flakes and agglomerated particles close to the surface and in the center of the droplet, respectively. The graphite morphologies, growth mechanisms, defects, irregularities and growth instabilities were interpreted with detailed optical and scanning electron microscopies.

Combined computational and experimental study of Ar beam induced defect formation in graphite

International Nuclear Information System (INIS)

Pregler, Sharon K.; Hayakawa, Tetsuichiro; Yasumatsu, Hisato; Kondow, Tamotsu; Sinnott, Susan B.

2007-01-01

Irradiation of graphite, commonly used in nuclear power plants, is known to produce structural damage. Here, experimental and computational methods are used to study defect formation in graphite during Ar irradiation at incident energies of 50 eV. The experimental samples are analyzed with scanning tunneling microscopy to quantify the size distribution of the defects that form. The computational approach is classical molecular dynamic simulations that illustrate the mechanisms by which the defects are produced. The results indicate that defects in graphite grow in concentrated areas and are nucleated by the presence of existing defects

Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

Energy Technology Data Exchange (ETDEWEB)

Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)

AGC 2 Irradiated Material Properties Analysis

Energy Technology Data Exchange (ETDEWEB)

Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

Theory of void swelling, irradiation creep and growth

International Nuclear Information System (INIS)

Wood, M.H.; Bullough, R.; Hayns, M.R.

Recent progress in our understanding of the fundamental mechanisms involved in swelling, creep and growth of materials subjected to irradiation is reviewed. The topics discussed are: the sink types and their strengths in the lossy continuum; swelling and void distribution analysis, including recent work on void nucleation; and, irradiation creep and growth of zirconium and zircaloy are taken as an example

The role of oxygen in the uptake of deuterium in lithiated graphite

Energy Technology Data Exchange (ETDEWEB)

Taylor, C. N.; Luitjohan, K. E. [School of Nuclear Engineering, Purdue University, West Lafayette, Indiana 47907 (United States); Dadras, J. [Department of Physics and Astronomy, University of Tennessee, Knoxville, Tennessee 37998 (United States); Allain, J. P. [School of Nuclear Engineering, Purdue University, West Lafayette, Indiana 47907 (United States); Birck Nanotechnology Center, West Lafayette, Indiana 47907 (United States); Krstic, P. S. [Department of Physics and Astronomy, University of Tennessee, Knoxville, Tennessee 37998 (United States); Joint Institute of Computational Sciences, University of Tennessee, Knoxville, Tennessee 37998 (United States); Physics Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Skinner, C. H. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2013-12-14

We investigate the mechanism of deuterium retention by lithiated graphite and its relationship to the oxygen concentration through surface sensitive experiments and atomistic simulations. Deposition of lithium on graphite yielded 5%–8% oxygen surface concentration and when subsequently irradiated with D ions at energies between 500 and 1000 eV/amu and fluences over 10{sup 16} cm{sup −2} the oxygen concentration rose to between 25% and 40%. These enhanced oxygen levels were reached in a few seconds compared to about 300 h when the lithiated graphite was allowed to adsorb oxygen from the ambient environment under equilibrium conditions. Irradiating graphite without lithium deposition, however, resulted in complete removal of oxygen to levels below the detection limit of XPS (e.g., <1%). These findings confirm the predictions of atomistic simulations, which had concluded that oxygen was the primary component for the enhanced hydrogen retention chemistry on the lithiated graphite surface.

Magnetic order in graphite: Experimental evidence, intrinsic and extrinsic difficulties

International Nuclear Information System (INIS)

Esquinazi, P.; Barzola-Quiquia, J.; Spemann, D.; Rothermel, M.; Ohldag, H.; Garcia, N.; Setzer, A.; Butz, T.

2010-01-01

We discuss recently obtained data using different experimental methods including magnetoresistance measurements that indicate the existence of metal-free high-temperature magnetic order in graphite. Intrinsic as well as extrinsic difficulties to trigger magnetic order by irradiation of graphite are discussed in view of recently published theoretical work.

Analysis of picosecond pulsed laser melted graphite

International Nuclear Information System (INIS)

Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

1986-01-01

A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm -1 and the disorder-induced mode at 1360 cm -1 , the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence

Graphite development for gas-cooled reactors in the USA

International Nuclear Information System (INIS)

Burchell, T.D.

1991-01-01

This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR

Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

International Nuclear Information System (INIS)

2014-01-01

Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

Energy Technology Data Exchange (ETDEWEB)

Eapen, Jacob [North Carolina State Univ., Raleigh, NC (United States); Murty, Korukonda [North Carolina State Univ., Raleigh, NC (United States); Burchell, Timothy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-06-02

Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

Elastic properties of graphite and interstitial defects

International Nuclear Information System (INIS)

Ayasse, J.-B.

1977-01-01

The graphite elastic constants C 33 and C 44 , reflecting the interaction of the graphitic planes, were experimentally measured as a function of irradiation and temperature. A model of non-central strength atomic interaction was established to explain the experimental results obtained. This model is valid at zero temperature. The temperature dependence of the elastic properties was analyzed. The influence of the elastic property variations on the specific heat of the lattice at very low temperature was investigated [fr

Depleted Hydrocarbon Reservoirs Present a Safe and Practical Burial Solution for Graphite Waste

International Nuclear Information System (INIS)

Rahmani, L.

2016-01-01

A solution for graphite waste is proposed that combines reliance on thick impermeable host rock that is needed to confine the long-life radioactivity content of most irradiated graphite with low capitalistic and operational unit volume costs that are required to render this bulky waste form manageable. The solution, uniquely applicable to irradiated graphite due to its low dose rates, moderate mechanical strength and light density, consists in three steps: first, graphite is fine-crushed under water; second, it is made in an aqueous suspension; third, the suspension is injected into a deep, disused hydrocarbon reservoir. Each of these steps only involves well mastered techniques. Regulatory changes that may allow this solution to be added to the gamut of available waste routes, geochemical issues, availability of depleted reservoirs and cost projections are presented. (author)

Effect of gamma radiation on graphite - PTFE dry lubrication system

Science.gov (United States)

Singh, Sachin; Tyagi, Mukti; Seshadri, Geetha; Tyagi, Ajay Kumar; Varshney, Lalit

2017-12-01

An effect of gamma radiation on lubrication behavior of graphite -PTFE dry lubrication system has been studied using (TR-TW-30L) tribometer with thrust washer attachment in plane contact. Different compositions of graphite and PTFE were prepared and irradiated by gamma rays. Gamma radiation exposure significantly improves the tribological properties indicated by decrease in coefficient of friction and wear properties of graphite -PTFE dry lubrication system. SEM and XRD analysis confirm the physico-chemical modification of graphite-PTFE on gamma radiation exposure leading to a novel dry lubrication system with good slip and anti friction properties.

AGC 2 Irradiation Creep Strain Data Analysis

Energy Technology Data Exchange (ETDEWEB)

Windes, William E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

AGC 2 Irradiation Creep Strain Data Analysis

International Nuclear Information System (INIS)

Windes, William E.; Rohrbaugh, David T.; Swank, W. David

2016-01-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

Differential expression of growth factors in irradiated mouse testes

International Nuclear Information System (INIS)

Mauduit, Claire; Siah, Ahmed; Foch, Marie; Chapet, Olivier; Clippe, Sebastien; Gerard, Jean-Pierre; Benahmed, Mohamed

2001-01-01

Purpose: By using as an experimental model the male mouse gonad, which contains both radiosensitive (germ) and radioresistant (somatic) cells, we have studied the growth factor (and/or receptor) expression of transforming growth factor-β receptor (TGFβ RI), stem cell factor (SCF), c-kit, Fas-L, Fas, tumor necrosis factor receptor (TNF R55), and leukemia inhibiting factor receptor (LIF-R) after local irradiation. Methods and Materials: Adult male mice were locally irradiated on the testes. Induction of apoptosis in the different testicular cell types following X-ray radiation was identified by the TdT-mediated dUTP Nick End Labeling (TUNEL) approach. Growth factor expression was evidenced by semiquantitative RT-PCR and Western blot analyses. Results: Apoptosis, identified through the TUNEL approach, occurred in radiosensitive testicular (premeotic) germ cells with the following kinetics: the number of apoptotic cells increased after 24 h (p<0.001) and was maximal 48 h after a 2-Gy ionizing radiation (p<0.001). Apoptotic cells were no longer observed 72 h after a 2-Gy irradiation. The number of apoptotic cells increased with the dose of irradiation (1-4 Gy). In the seminiferous tubules, the growth factor expression in premeiotic radiosensitive germ cells was modulated by irradiation. Indeed Fas, c-kit, and LIF-R expression, which occurs in (radiosensitive) germ cells, decreased 24 h after a 2-Gy irradiation, and the maximal decrease was observed with a 4-Gy irradiation. The decrease in Stra8 expression occurred earlier, at 4 h after a 2-Gy irradiation. In addition, a significant (p<0.03) decrease in Stra8 mRNA levels was observed at the lowest dose used (0.5 Gy, 48 h). Moreover, concerning a growth factor receptor, such as TGFβ RI, which is expressed both in radiosensitive and radioresistant cells, we observed a differential expression depending on the cell radiosensitivity after irradiation. Indeed, TGFβ RI expression was increased after irradiation in

Progress in radioactive graphite waste management

International Nuclear Information System (INIS)

2010-07-01

, especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

Some aspects of nuclear graphite production in France; Etude generale sur les graphites nucleaires produits en France

Energy Technology Data Exchange (ETDEWEB)

Gueron, J; Hering, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legendre, A [Pechiney, 75 - Paris (France)

1958-07-01

1) Manufacturing: A summary and results on the CEA-Pechiney purification process are given. Variations in the preparation of green pastes and their effects on graphitized material are described. 2) Physical and mechanical properties: Results are given on: - Statistics of dimensional variatior products having square cross-section. - Statistical variation of thermal expansion coefficients and of electrical conductivity. - Density of normals to carbon layer planes and their connexion with thermal expansion. - Stress-strain cycles and conclusions drawn therefrom. - Mechanical resistance and gas permeability of items for supporting fuel elements. 3) Behaviour under radiation: Alteration under radiation of French graphites irradiated either in G1 pile or in experimental piles, and thermal annealing of those alterations, are given. (author)Fren. [French] 1) Fabrication: On resume le procede d'epuration CEA-PECHINEY, ainsi que diverses modalites de preparation des pates et on expose les resultats obtenus. 2) Proprietes physiques et mecaniques: On indique le resultat d'etudes sur: - la statistique des dimensions de produits a section carree. - celle des variations des coefficients de dilatation thermique et de la conductibilite electrique. - la densite des normales aux plans graphitiques et leur connexion avec la dilatation thermique. - la compression mecanique du graphite. - la solidite mecanique et la permeabilite aux gaz de pieces destinees a supporter des cartouches de combustible. 3) Tenue sous rayonnement: Modification sous rayonnement des graphites fran is irradies soit dans la pile G1, soit dans des piles experimentales, et guerison thermique de ces modifications. (auteur)

IAEA International Database on Irradiated Nuclear Graphite Properties. 7th meeting of the Technical Steering Committee

International Nuclear Information System (INIS)

Humbert, D.; Wickham, A.J.

2005-06-01

This report summarizes the Consultant Meeting '7th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 March 2005 at the IAEA Headquarters, Vienna, Austria. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

IAEA International Database on Irradiated Nuclear Graphite Properties. 6th meeting of the Technical Steering Committee

International Nuclear Information System (INIS)

Humbert, D.; Wickham, A.J.

2004-12-01

This report summarizes the Consultant Meeting 6th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 September 2004 at Plas Tan-Y-Bwlch, Maentwrog, Gwynedd, UK. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

Contributions for the international conference on carbon and graphite CARBON '88

International Nuclear Information System (INIS)

Delle, W.

1988-08-01

This report is the compilation of three papers prepared by the Kernforschungsanlage Juelich GmbH (KFA) in collaboration with other partners for the International Conference CARBON '88. The topics were as follows: 1.) Fracture toughness of fast neutron irradiated graphite (W. Delle, H. Derz, G. Kleist, H. Nickel, W. Thiele); 2.) The irradiation creep characteristics of graphite to high fluences (C.R. Kennedy, M. Cundy, G. Kleist); and 3.) New silicon carbide materials starting with the Coat-Mix procedure (H.K. Luhleich, K. Bach, F.J. Dias, M. Kampel, F. Koch, H. Nickel). (orig./MM)

Trapping and detrapping of hydrogen in graphite materials exposed to hydrogen gas

International Nuclear Information System (INIS)

Atsumi, Hisao; Iseki, Michio; Shikama, Tatsuo.

1994-01-01

Measurements of hydrogen solubility have been performed for several unirradiated and neutron-irradiated graphite (and CFC) samples at temperatures between 973 and 1323 K under a ∼10 kPa hydrogen atmosphere. The hydrogen dissolution process has been studied and it is discussed here. The values of hydrogen solubility vary substantially among the samples up to about a factor of 16. A strong correlation has been observed between the values of hydrogen solubility and the degrees of graphitization determined by X-ray diffraction technique. The relation can be extended even for the neutron irradiated samples. Hydrogen dissolution into graphite can be explained with the trapping of hydrogen at defect sites (e.g. dangling carbon bonds) considering an equilibrium reaction between hydrogen molecules and the trapping sites. The migration of hydrogen in graphite is speculated to result from a sequence of detrapping and retrapping events with high energy activation processes. (author)

Increase of the density of commercial graphite

International Nuclear Information System (INIS)

Tobias, H.; Meyerstein, D.

1977-12-01

The increase of the density of commercial graphite of the type ATJ by polymerization of an impregnated monomer, followed by pyrolysis, is described. The monomer which was either styrene or acrylonitrile, was irradiated by a 60 Co source and pyrolized in a standard vacuum system. The irradiation dose for the polymerization of the monomer was determined. Suggestions for the establishment of the optimum conditions are offered

Brazing graphite to graphite

International Nuclear Information System (INIS)

Peterson, G.R.

1976-01-01

Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

Tritium retention properties of tungsten, graphite and co-deposited carbon film

International Nuclear Information System (INIS)

Nobuta, Y.; Hatano, Y.; Matsuyama, M.; Abe, S.; Akamaru, S.; Yamauchi, Y.; Hino, T.; Suzuki, S.; Akiba, M.

2014-01-01

DT + ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keV of DT + irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium

Self-propagating solar light reduction of graphite oxide in water

Energy Technology Data Exchange (ETDEWEB)

Todorova, N.; Giannakopoulou, T.; Boukos, N.; Vermisoglou, E. [Institute of Nanoscience and Nanotechnology, NCSR “Demokritos”, 153 41 Attikis (Greece); Lekakou, C. [Division of Mechanical, Medical, and Aerospace Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford (United Kingdom); Trapalis, C., E-mail: c.trapalis@inn.demokritos.gr [Institute of Nanoscience and Nanotechnology, NCSR “Demokritos”, 153 41 Attikis (Greece)

2017-01-01

Highlights: • Graphite oxide was partially reduced by solar light irradiation in water media. • No addition of catalysts nor reductive agent were used for the reduction. • Specific capacitance increased stepwise with increase of irradiation time. • Self-propagating reduction of graphene oxide by solar light is suggested. - Abstract: Graphite Oxide (GtO) is commonly used as an intermediate material for preparation of graphene in the form of reduced graphene oxide (rGO). Being a semiconductor with tunable band gap rGO is often coupled with various photocatalysts to enhance their visible light activity. The behavior of such rGO-based composites could be affected after prolonged exposure to solar light. In the present work, the alteration of the GtO properties under solar light irradiation is investigated. Water dispersions of GtO manufactured by oxidation of natural graphite via Hummers method were irradiated into solar light simulator for different periods of time without addition of catalysts or reductive agent. The FT-IR analysis of the treated dispersions revealed gradual reduction of the GtO with the increase of the irradiation time. The XRD, FT-IR and XPS analyses of the obtained solid materials confirmed the transition of GtO to rGO under solar light irradiation. The reduction of the GtO was also manifested by the CV measurements that revealed stepwise increase of the specific capacitance connected with the restoration of the sp{sup 2} domains. Photothermal self-propagating reduction of graphene oxide in aqueous media under solar light irradiation is suggested as a possible mechanism. The self-photoreduction of GtO utilizing solar light provides a green, sustainable route towards preparation of reduced graphene oxide. However, the instability of the GtO and partially reduced GO under irradiation should be considered when choosing the field of its application.

In-pile IASCC growth tests of irradiated stainless steels in JMTR

Energy Technology Data Exchange (ETDEWEB)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Shibata, Akira; Ohmi, Masao [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation-assisted stress corrosion cracking (IASCC) test plan to evaluate in-situ effects of neutron/{gamma}-ray irradiation on crack growth of irradiated stainless steels under high-temperature water conditions for commercial boiling water reactors (BWRs) using the Japan Materials Testing Reactor (JMTR). Crack growth rate and its electrochemical corrosion potential (ECP) dependence are different between in-pile test and post irradiation examination (PIE), but these differences are not fully understood. The objectives of the present study are to understand the difference between in-pile and out-of-pile IASCC growth and to confirm the effectiveness of mitigation due to lowering ECP on in-pile crack growth rates. For in-pile crack growth tests, we have selected a large compact tension specimen such as 0.5T-CT because of validity of SCC growth test at a high stress intensity factor (K-value). For loading a 0.5T-CT specimen up to K - 30 MPa {radical}m, we have adopted a lever type loading unit for in-pile crack growth tests in the JMTR. In this report, an in-pile test plan for crack growth of irradiated SUS316L stainless steels under simulated BWR conditions in the JMTR and current status of development of in-pile crack growth test techniques are presented. (author)

UV irradiation assisted growth of ZnO nanowires on optical fiber surface

Energy Technology Data Exchange (ETDEWEB)

Gong, Bo; Shi, Tielin; Liao, Guanglan; Li, Xiaoping; Huang, Jie; Zhou, Temgyuan; Tang, Zirong, E-mail: zirong@mail.hust.edu.cn

2017-06-01

Highlights: • A new fabrication process combined a hydrothermal process with UV irradiation from optical fiber is developed. • The growth of ZnO nanowires is efficient in the utilization of UV light. • A novel hybrid structure which integrates ZnO nanowires on optical fiber surface is synthesized. • The UV assisted growth of ZnO nanowires shows preferred orientation and better quality. • A mechanism of growing ZnO nanowires under UV irradiation is proposed. - Abstract: In this paper, a novel approach was developed for the enhanced growth of ZnO nanowires on optical fiber surface. The method combined a hydrothermal process with the efficient UV irradiation from the fiber core, and the effects of UV irradiation on the growth behavior of ZnO nanowires were investigated. The results show that UV irradiation had great effects on the preferred growth orientation and the quality of the ZnO nanowires. The crystallization velocity along the c-axis would increase rapidly with the increase of the irradiation power, while the growth process in the lateral direction was marginally affected by the irradiation. The structure of ZnO nanowires also shows less oxygen vacancy with UV irradiation of higher power. The developed approach is applicable for the efficient growth of nanowires on the fiber surface, and the ZnO nanowires/optical fiber hybrid structures have great potentials for a wide variety of applications such as optical fiber sensors and probes.

Modeling to study the role of catalyst in the formation of graphitic shells during carbon nanofiber growth subjected to reactive plasma

Science.gov (United States)

Gupta, Ravi; Gupta, Neha; Sharma, Suresh C.

2018-04-01

An analytical model to study the role of a metal catalyst nanofilm in the nucleation, growth, and resulting structure of carbon nanofibers (CNFs) in low-temperature hydrogen diluted acetylene plasma has been developed. The model incorporates the nanostructuring of thin catalyst films, growth of CNF, restructuring of catalyst nanoparticles during growth, and its repercussion on the resulting structure (alignment of rolled graphene sheets around catalyst nanoparticles) by taking into account the plasma sheath formalization, kinetics of neutrals and positively charged species in the reactive plasma, flux of plasma species onto the catalyst front surface, and numerous surface reactions for carbon generation. In order to examine the influence of the catalyst film on the growth of CNFs, the numerical solutions of the model equations have been obtained for experimentally determined initial conditions and glow discharge plasma parameters. From the solutions obtained, we found that nanostructuring of thin films leads to the formation of small nanoparticles with high surface number density. The CNF nucleates over these small-sized nanoparticles grow faster and attain early saturation because of the quick poisoning of small-sized catalyst particles, and contain only a few graphitic shells. However, thick nanofilms result in shorter CNFs with large diameters composed of many graphitic shells. Moreover, we found that the inclination of graphitic shells also depends on the extent up to which the catalyst can reconstruct itself during the growth. The small nanoparticles show much greater elongation along the growth axis and also show a very small difference between their tip and base diameter during the growth due to which graphitic shells align at very small angles as compared to the larger nanoparticles. The present study is useful to synthesize the thin and more extended CNFs/CNTs having a smaller opening angle (inclination angle of graphene layers) as the opening angle has a

Irradiation creep due to SIPA-induced growth

International Nuclear Information System (INIS)

Woo, C.H.

1980-01-01

An additional contribution to irradiation creep resulting from the stress-induced preferred adsorption (SIPA) effect is described - SIPA-induced growth (SIG). The mechanism of SIG is discussed and an expression for its contribution to irradiation creep developed. It is shown that SIG is very significant in comparison with SIPA. Enhancement of creep by swelling may also occur. (U.K.)

Irradiation test HT-31: high-temperature irradiation behavior of LASL-made extruded fuel rods and LASL-made coated particles

International Nuclear Information System (INIS)

Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; Davidson, K.V.; Schell, D.H.

1977-04-01

Three LASL-made extruded graphite and coated particle fuel rods have been irradiated in the Oak Ridge National Laboratory High Fluence Isotope Reactor test HT-31. Test conditions were about 9 x 10 21 nvt(E > .18 MeV) at 1250 0 C. The graphite matrix showed little or no effect of the irradiation. LASL-made ZrC containing coated particles with ZrC coats and ZrC-doped pyrolytic carbon coats showed no observable effects of the irradiation

γ-Irradiation assisted synthesis of graphene oxide sheets supported Ag nanoparticles with single crystalline structure and parabolic distribution from interlamellar limitation

Energy Technology Data Exchange (ETDEWEB)

Yue, Yunhao; Zhou, Baoming; Shi, Jie; Chen, Cheng; Li, Nan; Xu, Zhiwei, E-mail: xuzhiwei@tjpu.edu.cn; Liu, Liangsen; Kuang, Liyun; Ma, Meijun; Fu, Hongjun

2017-05-01

Highlights: • Graphene oxide sheets supported Ag nanoparticles composites are successfully prepared via γ-irradiation without surfactant or functional agent. • Ag nanoparticles exhibit single crystalline structure and parabolic distribution on the surface of graphene oxide sheets. • Proposing a view that the growth of intercellular AgNPs can be limited by graphite oxide. - Abstract: This paper reported a method to fabricate graphene oxide sheets supported Ag nanoparticles (AgNPs/GOS) with single crystalline structure and parabolic distribution without surfactant or functional agent. We used imidazole silver nitrate as intercalation precursor into the layers of graphite oxide, and subsequently reduction and growth of interlamellar AgNPs were induced via γ-irradiation. The results illustrated that the synergism of interlamellar limitation of graphite oxide and fragmentation ability of γ-irradiation could prevent coalescent reaction of AgNPs with other oligomeric clusters, and the single crystalline and small-sized (below 13.9 nm) AgNPs were prepared. Moreover, the content and size of AgNPs exhibited parabolic distribution on GOS surface because the graphite oxide exfoliated to GOS from the edge to the central area of layers. In addition, complete exfoliation degree of GOS and large-sized AgNPs were obtained simultaneously under suitable silver ions concentration. Optimized composites exhibited outstanding surface-enhanced Raman scattering properties for crystal violet with enhancement factor of 1.3 × 10{sup 6} and detection limit of 1.0 × 10{sup −7} M, indicating that the AgNPs/GOS composites could be applied to trace detection of organic dyes molecules. Therefore, this study presented a strategy for developing GOS supported nanometal with single crystalline structure and parabolic distribution based on γ-irradiation.

γ-Irradiation assisted synthesis of graphene oxide sheets supported Ag nanoparticles with single crystalline structure and parabolic distribution from interlamellar limitation

International Nuclear Information System (INIS)

Yue, Yunhao; Zhou, Baoming; Shi, Jie; Chen, Cheng; Li, Nan; Xu, Zhiwei; Liu, Liangsen; Kuang, Liyun; Ma, Meijun; Fu, Hongjun

2017-01-01

Highlights: • Graphene oxide sheets supported Ag nanoparticles composites are successfully prepared via γ-irradiation without surfactant or functional agent. • Ag nanoparticles exhibit single crystalline structure and parabolic distribution on the surface of graphene oxide sheets. • Proposing a view that the growth of intercellular AgNPs can be limited by graphite oxide. - Abstract: This paper reported a method to fabricate graphene oxide sheets supported Ag nanoparticles (AgNPs/GOS) with single crystalline structure and parabolic distribution without surfactant or functional agent. We used imidazole silver nitrate as intercalation precursor into the layers of graphite oxide, and subsequently reduction and growth of interlamellar AgNPs were induced via γ-irradiation. The results illustrated that the synergism of interlamellar limitation of graphite oxide and fragmentation ability of γ-irradiation could prevent coalescent reaction of AgNPs with other oligomeric clusters, and the single crystalline and small-sized (below 13.9 nm) AgNPs were prepared. Moreover, the content and size of AgNPs exhibited parabolic distribution on GOS surface because the graphite oxide exfoliated to GOS from the edge to the central area of layers. In addition, complete exfoliation degree of GOS and large-sized AgNPs were obtained simultaneously under suitable silver ions concentration. Optimized composites exhibited outstanding surface-enhanced Raman scattering properties for crystal violet with enhancement factor of 1.3 × 10"6 and detection limit of 1.0 × 10"−"7 M, indicating that the AgNPs/GOS composites could be applied to trace detection of organic dyes molecules. Therefore, this study presented a strategy for developing GOS supported nanometal with single crystalline structure and parabolic distribution based on γ-irradiation.

Single-crystal apatite nanowires sheathed in graphitic shells: synthesis, characterization, and application.

Science.gov (United States)

Jeong, Namjo; Cha, Misun; Park, Yun Chang; Lee, Kyung Mee; Lee, Jae Hyup; Park, Byong Chon; Lee, Junghoon

2013-07-23

Vertically aligned one-dimensional hybrid structures, which are composed of apatite and graphitic structures, can be beneficial for orthopedic applications. However, they are difficult to generate using the current method. Here, we report the first synthesis of a single-crystal apatite nanowire encapsulated in graphitic shells by a one-step chemical vapor deposition. Incipient nucleation of apatite and its subsequent transformation to an oriented crystal are directed by derived gaseous phosphorine. Longitudinal growth of the oriented apatite crystal is achieved by a vapor-solid growth mechanism, whereas lateral growth is suppressed by the graphitic layers formed through arrangement of the derived aromatic hydrocarbon molecules. We show that this unusual combination of the apatite crystal and the graphitic shells can lead to an excellent osteogenic differentiation and bony fusion through a programmed smart behavior. For instance, the graphitic shells are degraded after the initial cell growth promoted by the graphitic nanostructures, and the cells continue proliferation on the bare apatite nanowires. Furthermore, a bending experiment indicates that such core-shell nanowires exhibited a superior bending stiffness compared to single-crystal apatite nanowires without graphitic shells. The results suggest a new strategy and direction for bone grafting materials with a highly controllable morphology and material conditions that can best stimulate bone cell differentiation and growth.

Non-activated high surface area expanded graphite oxide for supercapacitors

Energy Technology Data Exchange (ETDEWEB)

Vermisoglou, E.C.; Giannakopoulou, T.; Romanos, G.E.; Boukos, N.; Giannouri, M. [Institute of Nanoscience and Nanotechnology “Demokritos”, 153 43 Ag. Paraskevi, Attikis (Greece); Lei, C.; Lekakou, C. [Division of Mechanical, Medical, and Aerospace Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford GU2 7XH (United Kingdom); Trapalis, C., E-mail: c.trapalis@inn.demokritos.gr [Institute of Nanoscience and Nanotechnology “Demokritos”, 153 43 Ag. Paraskevi, Attikis (Greece)

2015-12-15

Graphical abstract: - Highlights: • One-step exfoliation and reduction of graphite oxide via microwave irradiation. • Effect of pristine graphite (type, flake size) on the microwave expanded material. • Effect of pretreatment and oxidation cycles on the produced expanded material. • Expanded graphene materials with high BET surface areas (940 m{sup 2}/g–2490 m{sup 2}/g). • Non-activated graphene based materials suitable for supercapacitors. - Abstract: Microwave irradiation of graphite oxide constitutes a facile route toward production of reduced graphene oxide, since during this treatment both exfoliation and reduction of graphite oxide occurs. In this work, the effect of pristine graphite (type, size of flakes), pretreatment and oxidation cycles on the finally produced expanded material was examined. All the types of graphite that were tested afforded materials with high BET surface areas ranging from 940 m{sup 2}/g to 2490 m{sup 2}/g, without intervening an activation stage at elevated temperature. SEM and TEM images displayed exfoliated structures, where the flakes were significantly detached and curved. The quality of the reduced graphene oxide sheets was evidenced both by X-ray photoelectron spectroscopy and Raman spectroscopy. The electrode material capacitance was determined via electrochemical impedance spectroscopy and cyclic voltammetry. The materials with PEDOT binder had better performance (∼97 F/g) at low operation rates while those with PVDF binder performed better (∼20 F/g) at higher rates, opening up perspectives for their application in supercapacitors.

A study of the coefficient of thermal expansion of nuclear graphites

International Nuclear Information System (INIS)

Hacker, P.J.

2001-02-01

This thesis presents the results of a study of the Coefficient of Thermal Expansion (CTE) of two grades of nuclear graphite that are used as the moderator in the Magnox and Advanced Gas-Cooled reactors operated in the UK. This work has two main aims, the first is to characterise those elements of the graphite microstructure that control CTE within these materials and to relate these to the effects induced within the reactor. The second is to develop a microstructural model, of general applicability, that can initially be applied to model the CTE changes within the graphites under reactor conditions (neutron irradiation and radiolytic oxidation). These aims have been met by study in three interlinked areas, theoretical, experimental and modelling. Previous to this study, a loose assembly of single crystals together with changes in small scale nanometric porosity (Mrozowski cracks) were used to describe CTE behaviour of nuclear graphite both as-received and under reactor conditions. Within the experimental part of this thesis the graphite nanostructure was studied using, primarily, Transmission Electron Microscopy (TEM). This work concluded that structure on this scale was complex and that the loose assembly of single crystals was a poor microstructural approximation for modelling the CTE of these materials. Other experimental programmes measured the CTE of highly oxidised samples and simulated the effects of irradiation. The former discovered that CTE remained largely unaffected to high weight losses. This insensitivity was explained by ''The Continuous Network Hypothesis'' that was also related to classical percolation theory. The final part of the thesis modelled an abstraction of the key microstructural features identified in the previous parts of the thesis. This approach has been applied to AGR moderator graphite where it has successfully modelled the thermal expansion behaviour of the as-received, irradiated and oxidised material. (author)

Characterization of radiation damage induced by swift heavy ions in graphite

Energy Technology Data Exchange (ETDEWEB)

Hubert, Christian

2016-05-15

Graphite is a classical material in neutron radiation environments, being widely used in nuclear reactors and power plants as a moderator. For high energy particle accelerators, graphite provides ideal material properties because of the low Z of carbon and its corresponding low stopping power, thus when ion projectiles interact with graphite is the energy deposition rather low. This work aims to improve the understanding of how the irradiation with swift heavy ions (SHI) of kinetic energies in the range of MeV to GeV affects the structure of graphite and other carbon-based materials. Special focus of this project is given to beam induced changes of thermo-mechanical properties. For this purpose the Highly oriented pyrolytic graphite (HOPG) and glassy carbon (GC) (both serving as model materials), isotropic high density polycrystalline graphite (PG) and other carbon based materials like carbon fiber carbon composites (CFC), chemically expanded graphite (FG) and molybdenum carbide enhanced graphite composites (MoC) were exposed to different ions ranging from {sup 131}Xe to {sup 238}U provided by the UNILAC accelerator at GSI in Darmstadt, Germany. To investigate structural changes, various in-situ and off-line measurements were performed including Raman spectroscopy, x-ray diffraction and x-ray photo-electron spectroscopy. Thermo-mechanical properties were investigated using the laser-flash-analysis method, differential scanning calorimetry, micro/nano-indentation and 4-point electrical resistivity measurements. Beam induced stresses were investigated using profilometry. Obtained results provided clear evidence that ion beam-induced radiation damage leads to structural changes and degradation of thermal, mechanical and electrical properties of graphite. PG transforms towards a disordered sp2 structure, comparable to GC at high fluences. Irradiation-induced embrittlement is strongly reducing the lifetime of most high-dose exposed accelerator components. For

Thermal migration of deuterium implanted in graphite: Influence of free surface proximity and structure

Energy Technology Data Exchange (ETDEWEB)

Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Moncoffre, N., E-mail: n.moncoffre@ipnl.in2p3.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Toulhoat, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Ammar, M.R. [CNRS, CEMHTI UPR3079, Université Orléans, CS90055, F-45071 Orléans cedex 2 (France); Rouzaud, J.N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure, Paris, UMR CNRS ENS 8538, F-75231 Paris cedex 5 (France)

2016-03-15

This paper is a contribution to the study of the behavior of activation products produced in irradiated nuclear graphite, graphite being the moderator of the first French generation of CO{sub 2} cooled nuclear fission reactors. This paper is focused on the thermal release of Tritium, a major contributor to the initial activity, taking into account the role of the free surfaces (open pores and graphite surface). Two kinds of graphite were compared. On one hand, Highly Oriented Pyrolitic Graphite (HOPG), a model well graphitized graphite, and on the other hand, SLA2, a porous less graphitized nuclear graphite. Deuterium ion implantation at three different energies 70, 200 and 390 keV allows simulating the presence of Tritium at three different depths, corresponding respectively to projected ranges R{sub p} of 0.75, 1.7 and 3.2 μm. The D isotopic tracing is performed thanks to the D({sup 3}He,p){sup 4}He nuclear reaction. The graphite structure is studied by Raman microspectrometry. Thermal annealing is performed in the temperature range 200–1200 °C up to 300 h annealing time. As observed in a previous study, the results show that the D release occurs according to three kinetic regimes: a rapid permeation through open pores, a transient regime corresponding to detrapping and diffusion of D located at low energy sites correlated to the edges of crystallites and finally a saturation regime attributed to detrapping of interstitial D located at high energy sites inside the crystallites. Below 600 °C, D release is negligible whatever the implantation depth and the graphite type. The present paper clearly puts forward that above 600 °C, the D release decreases at deeper implantation depths and strongly depends on the graphite structure. In HOPG where high energy sites are more abundant, the D release is less dependent on the surface proximity compared to SLA2. In SLA2, in which the low energy sites prevail, the D release curves are clearly shifted towards lower

A comparative study of post-irradiation growth kinetics of spheroids and monolayers

International Nuclear Information System (INIS)

Dertinger, J.; Luecke-Huhle, C.

1975-01-01

Post-irradiation growth kinetics of γ-irradiated spheroid and monolayer cells in exponential growth phase was investigated by means of dose-response curves based on cell counts after specified time intervals following irradiation. A mathematical model of cell-growth after irradiation was fitted to these curves. The model parameters (related to division delay and growth of non-surviving cells) obtained from this analysis consistently indicated increasing resistance to sub-lethal damage of cells cultured as multicellular spheroids under conditions of increasing three-dimensional contact. In contrast, no indication of an increased radiation-resistance was found with cells cultured on a substratum under a variety of conditions. (author)

Investigation of carbon near the graphite-diamond-liquid triple point

International Nuclear Information System (INIS)

Prawer, S.; Jamieson, D.N.

1992-01-01

Pulsed laser irradiation is used to heat deeply buried damage layers in diamond. Over a small range of laser powers, damage annealing, formation of buried graphitic layers, and melting of diamond followed by its conversion upon cooling into graphite are observed. The diagnostics employed are Channeling Contrast Microscopy, optical absorption, surface profilometry, and scanning and optical microscopies. The results are explained in terms of the behaviour of carbon under high internal pressures close to the diamond-graphite-liquid carbon triple point in the phase diagram. 17 refs., 3 figs

Analysis of the microbial growth in 60Co γ-irradiated foods by calorimetry

International Nuclear Information System (INIS)

Furuta, Masakazu; Hayashi, Toshio; Hamasaki, Koji; Wirkner, Sandra; Constantinoiu, Elena; Takahashi, Katsutada

2002-01-01

Using a heat conduction calorimeter equipped with 24 sample units the heat evolutions from growing 60 Co γ-irradiated bioburden of black pepper seeds and frozen beef were detected in the form of growth thermograms. 60 Co γ-irradiation affected the growth pattern in which a dose-dependent reduction of the growth rate constant was observed together with the retardation in growth, indicating a combination of bactericidal and bacteriostatic effects. We successfully determined the minimal inactivation doses for the two food samples using the relationship between the irradiation dose and the retardation in growth t α , or the growth rate constant μ obtained from the growth thermograms. These results strongly suggested the possibility of calorimetry as measure of predictive microbiology in food irradiation. (author)

Analysis of the microbial growth in 60Co gamma-irradiated foods by calorimetry

International Nuclear Information System (INIS)

Furuta, M.; Hamasaki, K.; Wirkner, S.; Constantinoiu, E.; Takahashi, K.; Hayashi, T.

2002-01-01

Using a heat conduction calorimeter equipped with 24 sample units the heat evolutions from growing 60Co gamma-irradiated bioburden of black pepper seeds and frozen beef were detected in the form of growth thermograms. 60Co gamma-irradiation affected the growth pattern in which a dose-dependent reduction of the growth rate constant was observed together with the retardation in growth, indicating a combination of bactericidal and bacteriostatic effects. We successfully determined the minimal inactivation doses for the two food samples using the relationship between the irradiation dose and the retardation in growth talpha, or the growth rate constant mu obtained from the growth thermograms. These results strongly suggested the possibility of calorimetry as a measure of predictive microbiology in food irradiation

Preparation of graphite dispersed copper composite on copper plate with CO2 laser

Science.gov (United States)

Yokoyama, S.; Ishikawa, Y.; Muizz, M. N. A.; Hisyamudin, M. N. N.; Nishiyama, K.; Sasano, J.; Izaki, M.

2018-01-01

It was tried in this work to prepare the graphite dispersed copper composite locally on a copper plate with a CO2 laser. The objectives of this study were to clear whether copper graphite composite was prepared on a copper plate and how the composite was prepared. The carbon content at the laser spot decreased with the laser irradiation time. This mainly resulted from the elimination by the laser trapping. The carbon content at the outside of the laser spot increased with time. Both the laser ablation and the laser trapping did not act on the graphite particles at the outside of the laser spot. Because the copper at the outside of the laser spot melted by the heat conduction from the laser spot, the particles were fixed by the wetting. However, the graphite particles were half-floated on the copper plate. The Vickers hardness decreased with an increase with laser irradiation time because of annealing.

Direct synthesis of graphitic mesoporous carbon from green phenolic resins exposed to subsequent UV and IR laser irradiations

Science.gov (United States)

Sopronyi, Mihai; Sima, Felix; Vaulot, Cyril; Delmotte, Luc; Bahouka, Armel; Matei Ghimbeu, Camelia

2016-01-01

The design of mesoporous carbon materials with controlled textural and structural features by rapid, cost-effective and eco-friendly means is highly demanded for many fields of applications. We report herein on the fast and tailored synthesis of mesoporous carbon by UV and IR laser assisted irradiations of a solution consisting of green phenolic resins and surfactant agent. By tailoring the UV laser parameters such as energy, pulse repetition rate or exposure time carbon materials with different pore size, architecture and wall thickness were obtained. By increasing irradiation dose, the mesopore size diminishes in the favor of wall thickness while the morphology shifts from worm-like to an ordered hexagonal one. This was related to the intensification of phenolic resin cross-linking which induces the reduction of H-bonding with the template as highlighted by 13C and 1H NMR. In addition, mesoporous carbon with graphitic structure was obtained by IR laser irradiation at room temperature and in very short time periods compared to the classical long thermal treatment at very high temperatures. Therefore, the carbon texture and structure can be tuned only by playing with laser parameters, without extra chemicals, as usually required. PMID:28000781

Structural performance of a graphite blanket in fusion reactors

International Nuclear Information System (INIS)

Wolfer, W.G.; Watson, R.D.

1978-01-01

Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

Calculation of thermal stresses in graphite fuel blocks

International Nuclear Information System (INIS)

Lejeail, Y.; Cabrillat, M.T.

2005-01-01

This paper presents a parametric study of temperature and thermal stress calculations inside a HTGR core graphite block, taking into account the effect of fluence on the thermal and mechanical properties, up to 4. 10 21 n/cm 2 . The Finite Element model, realized with Cast3M CEA code, includes the effects of irradiation creep, which tends to produce secondary stress relaxation. Then, the Weibull weakest link theory is recalled, evaluating the possible effects of volume, stress field distribution (loading factor), and multiaxiality for graphite-type materials, and giving the methodology to compare the stress to rupture for the structure to the one obtained from characterization, in the general case. The maximum of the Weibull stress in Finite Element calculations is compared to the value for tensile specimens. It is found that the maximum of the stress corresponds to the end of the irradiation cycle, after reactor shutdown, since both thermal conductivity and Young's modulus increase with time. However, this behaviour is partly counterbalanced by the increase of material strength with irradiation. (authors)

Radiation creep of graphite. An introduction

Energy Technology Data Exchange (ETDEWEB)

Blackstone, R [Commission of the European Communities, Petten (Netherlands). Joint Nuclear Research Center

1977-03-01

Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted.

Radiation creep of graphite. An introduction

International Nuclear Information System (INIS)

Blackstone, R.

1977-01-01

Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behavior compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted

Characterization of graphite dust produced by pneumatic lift

Energy Technology Data Exchange (ETDEWEB)

Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

2016-08-15

Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

A 3-D inelastic analysis of HTR graphite structures and a comparison with A 2-D approach

International Nuclear Information System (INIS)

Willaschek, J.

1979-01-01

In High Temperature Reactor Cores (HTR) a large number of elements are constructed of nuclear graphite. The dimensions of the graphite components are limited by stresses and strains resulting from thermal loads, irradiation induced dimensional changes and stress-dependent irradiation creep. Therefore it is necessary to examine the feasibility of design concepts with regard to the structural integrity of the material. This paper presents an analysis of a radial reflector concept for use in a 3000 MWth HTR for process heat production. This concept of a pebble bed reactor (OTTO cycle) requires reflector dimensions and shapes which have previously not been used and which may exceed acceptable stress limits. Graphite reflector elements in a HTR are subject to a high fluence of fast neutrons. The fluence varies spatially within an element. Irradiation-induced strains occur which in turn vary non-linearly with the fluence. At low fluences the graphite shrinks. With increasing fluence shrinkage is saturated and after a 'turn-around' point the graphite begins to swell. The net effect of fluence gradient and irradiation-induced strain is a 'necking' of the element which moves radially outwards with time. In this paper a three-dimensional inelastic analysis of a graphite block with the above deformation history is described. The influence of irradiation on dimensional stability and other material properties was taken into account. Numerical results were obtained with the finite-element computer code ADINA, modified at INTERATOM for the task in hand. The radial reflector block was modelled using 21-node three-dimensional continuum elements of elastic-creep material. The element stiffness matrices were calculated using the standard 2x2x2 Gauss integration; material nonlinearities with quadratic displacement functions and linearised initial strains were employed. (orig.)

Effect of gamma radiation on graphite – PTFE dry lubrication system

International Nuclear Information System (INIS)

Singh, Sachin; Tyagi, Mukti; Seshadri, Geetha; Tyagi, Ajay Kumar; Varshney, Lalit

2017-01-01

An effect of gamma radiation on lubrication behavior of graphite -PTFE dry lubrication system has been studied using (TR-TW-30L) tribometer with thrust washer attachment in plane contact. Different compositions of graphite and PTFE were prepared and irradiated by gamma rays. Gamma radiation exposure significantly improves the tribological properties indicated by decrease in coefficient of friction and wear properties of graphite -PTFE dry lubrication system. SEM and XRD analysis confirm the physico-chemical modification of graphite-PTFE on gamma radiation exposure leading to a novel dry lubrication system with good slip and anti friction properties. - Highlights: • Novel dry lubrication system of graphite -PTFE using gamma radiation. • Gamma radiation processing. • Reduction in coefficient of friction, frictional torque and wear loss of developed dry lubrication system.

Irradiation creep and growth behavior of Zircaloy-4 inner shell of HANARO

Energy Technology Data Exchange (ETDEWEB)

Chung, Jong-Ha; Cho, Yeong-Garp; Kim, Jong-In [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

2012-03-15

The inner shell of the reflector vessel of HANARO was made of Zircaloy-4 rolled plate. Zircaloy-4 rolled plate shows highly anisotropic behavior by fast neutron irradiation. This paper describes the analysis method for the irradiation induced creep and growth of the inner shell of HANARO. The anisotropic irradiation creep behavior was modeled as uniaxial strain-hardening power law modified by Hill's stress potential and the anisotropic irradiation growth was modeled by using volumetric swelling with anisotropic strain rate. In this study, the irradiation induced creep and growth behavior of the inner shell of the HANARO reflector vessel was re-evaluated. The rolling direction, the fast neutron flux, and the boundary conditions were applied with the same conditions as the actual inner shell. Analysis results show that deformation of the inner shell due to irradiation does not raise any problem for the lifetime of HANARO. (author)

Theoretical basis for graphite stress analysis in BERSAFE

International Nuclear Information System (INIS)

Harper, P.G.

1980-03-01

The BERSAFE finite element computer program for structural analysis has been extended to deal with structures made from irradiated graphite. This report describes the material behaviour which has been modelled and gives the theoretical basis for the solution procedure. (author)

Ag-catalyzed InAs nanowires grown on transferable graphite flakes

DEFF Research Database (Denmark)

Meyer-Holdt, Jakob; Kanne, Thomas; Sestoft, Joachim E.

2016-01-01

on exfoliated graphite flakes by molecular beam epitaxy. Ag catalyzes the InAs nanowire growth selectively on the graphite flakes and not on the underlying InAs substrates. This allows for easy transfer of the flexible graphite flakes with as-grown nanowire ensembles to arbitrary substrates by a micro...

Progress in radioactive graphite waste management. Additional information

International Nuclear Information System (INIS)

2010-06-01

, especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

Nanostructured carbon films with oriented graphitic planes

International Nuclear Information System (INIS)

Teo, E. H. T.; Kalish, R.; Kulik, J.; Kauffmann, Y.; Lifshitz, Y.

2011-01-01

Nanostructured carbon films with oriented graphitic planes can be deposited by applying energetic carbon bombardment. The present work shows the possibility of structuring graphitic planes perpendicular to the substrate in following two distinct ways: (i) applying sufficiently large carbon energies for deposition at room temperature (E>10 keV), (ii) utilizing much lower energies for deposition at elevated substrate temperatures (T>200 deg. C). High resolution transmission electron microscopy is used to probe the graphitic planes. The alignment achieved at elevated temperatures does not depend on the deposition angle. The data provides insight into the mechanisms leading to the growth of oriented graphitic planes under different conditions.

Mechanisms of irradiation growth of alpha-zirconium alloys

International Nuclear Information System (INIS)

Holt, R.A.

1988-01-01

Experimental observations in the last few years have shown that the range of irradiation growth behaviour of alpha-zirconium alloys is more varied, that a wider variety of sinks must be considered, and that there are more potential sources of anisotropy than was previously recognized. The important new experimental observations which influence our preception of the growth phenomenon in zirconium alloys include the growth of single crystals, accelerating growth in annealed material with the coincident appearance of vacancy loops on the basal planes, the occurrence of 'negative' growth, i.e., contractions along prism directions, the absence of a pronounced effect of grain size on the long term growth rate at low temperatures, and the presence of intergranular constraints prior to irradiation. With the greater complexity of behaviour now being observed, it is necessary to apply new theoretical concepts to assist in understanding growth, e.g., the potential role of anisotropic diffusion in segregation point defects to different sinks and 'growth' caused by the anisotropic relaxation of intergranular constrains. These can be combined with earlier ideas to predict a variety of growth behaviours, including 'negative growth'. Because the most important physical information required for theoretical treatments of growth, i.e, the characteristics of vacancies and self interstitial atoms, are still poorly understood, it is almost impossible to test rigorously any particular theoretical concept and a complete picture of growth has yet to emerge. (orig./MM)

Effects of pre-irradiation on isogeneic and semi-isogeneic CFU growth

International Nuclear Information System (INIS)

Buurman, W.A.; Bruggen, I. Van.

1977-01-01

The genetic resistance to a parental bone marrow transplant as demonstrated, when transplantation were performed early after irradiation, failed to occur if the interval between irradiation and transplantation was increased to 4 days. A similar radiation induced weakening of genetic resistance to a parental bone marrow graft in spleen and bone marrow would be demonstrated in mice, which has been irradiated with a sublethal dose at 7 days prior to the lethal irradiation and transplantation. The pre-irradiation of the recipient with a sublethal dose induced an enhancement of the growth in spleen and bone marrow of isogenic transplanted CFU. The pre-irradiation of a single tibia also resulted in a significant weakening of the resistance in the spleen. The experiments with partial body pre-irradiation suggested a local effect of the pre-irradiation, but it could be shown that the enhanced CFU growth is not caused by an enhanced seeding of CFU in pre-irradiation bone marrow. The role of microenvironment in the phenomenon of genetic resistance is discussed. (author)

The relationship between irradiation induced dimensional change and the coefficient of thermal expansion: A new look

International Nuclear Information System (INIS)

Hall, G.; Marsden, B.J.; Fok, A.; Smart, J.

2002-01-01

In the 1960s, J.H.W. Simmons derived a theoretical relationship between the coefficient of thermal expansion (CTE) and dimensional changes in irradiated graphite. At low irradiation dose, the theory was shown to be consistent with experimental observations. However, at higher doses the results diverge. Despite this, modified versions of this theory have been used as the basis of the design and life prediction calculations for graphite-moderated reactors. This paper revisits Simmons's theory, summarising the assumptions made in its derivation. The paper then modifies and applies the theory to the dimensional change and CTE change behaviour in isotropic nuclear graphite, making use of trends in irradiated behaviour recently derived using finite element analyses. The importance of these issues to present HTR technology is that the life of HTR graphite components is related to their irradiated dimensional change behaviour. A more in depth understanding of this behaviour will allow suitable graphite material to be selected or new graphite types to be developed. (author)

A TEM Study on the Microstructure of Fine Flaky Graphite

DEFF Research Database (Denmark)

Moumeni, Elham; Tiedje, Niels Skat; Horsewell, Andy

In this investigation the microstructure of the graphite flakes in titanium alloyed cast iron is studied using electron microscopy techniques. Based on this information, growth models for the platelets in the fine graphite flakes in cast iron are considered. Detailed crystallographic analysis...... of the defects observed such as multiple twin boundaries and possible spiral growth configurations are required...

Testing of Small Graphite Samples for Nuclear Qualification

Energy Technology Data Exchange (ETDEWEB)

Julie Chapman

2010-11-01

Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

Structural changes induced by electron irradiation

International Nuclear Information System (INIS)

Koike, J.; Pedraza, D.F.

1993-01-01

Highly oriented pyrolytic graphite was irradiated at room temperature with 300 kV electrons. Transmission electron microscopy and electron energy loss spectroscopy were employed to study the structural changes produced by irradiation. The occurrence of a continuous ring intensity in the selected area diffraction (SAD) pattern obtained on a specimen irradiated with the electron beam parallel to the c-crystallographic axis indicated that microstructural changes had occurred. However, from the SAD pattern obtained for the specimens tilted relative to the irradiation direction, it was found that up to a fluence of 1.1x10 27 e/m 2 graphite remained crystalline. An SAD pattern of a specimen irradiated with the electron beam perpendicular to the c-axis confirmed the persistence of crystalline order. High resolution electron microscopy showed that ordering along the c-axis direction remained. A density reduction of 8.9% due to irradiation was determined from the plasmon frequency shift. A qualitative model is proposed to explain these observations. A new determination of the threshold displacement energy, Ed, of carbon atoms in graphite was done by examining the appearance of a continuous ring in the SAD pattern at various electron energies. A value of 30 eV was obtained whether the incident electron beam was parallel or perpendicular to the c-axis, demonstrating that Ed is independent of the displacement direction

Development and engineering plan for graphite spent fuels conditioning program

International Nuclear Information System (INIS)

Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

1993-09-01

Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

Radiation creep of graphite. An introduction

International Nuclear Information System (INIS)

Blackstone, R.

1977-01-01

Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted. (Auth.)

The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

International Nuclear Information System (INIS)

Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

2014-01-01

Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

Effect of microstructure on helium bubble growth in irradiated nickel

International Nuclear Information System (INIS)

Sattler, M.L.

1986-01-01

Thin nickel films were irradiated with 80 keV helium ions at varying doses and varying temperatures in order to obtain a variety of final microstructures. The growth of bubbles was examined during in-situ irradiations at 950 0 C where migration and coalescence events were observed for bubbles as large as 60 nm. Further direct observations of bubble growth were made during annealing of the irradiated specimens. For sample with no visible bubbles before annealing, the heating to 0.51 T/sub M/ produced bubbles that increased in diameter with annealing time to the power n. For bubbles in the grain interior, n ∼ 1, and on the grain boundaries, n ∼ 0.6. Since no migration and coalescence or ripening theories predict this behavior, a theory described by transient diffusion to spherical sinks was developed to discuss the behavior. This theory predicts that n = 1 for bubbles growing in the grain interior and n = 0.5 for bubbles on the grain boundary. In other annealing of irradiated samples containing large bubble populations, the growth of large bubbles and shrinking of small bubbles was observed at a temperature equal to 0.54 T/sub M/. The theory of Ostwald ripening properly described this type of bubble growth. Mass spectrometer measurements of He content in the irradiated specimens showed a greater He retention in the Ni films that contained a significant bubble population than those with no visible bubbles

On the Crystallization of Compacted and Chunky Graphite from Liquid Multicomponent Iron-Carbon-Silicon-Based Melts

Science.gov (United States)

Stefanescu, D. M.; Huff, R.; Alonso, G.; Larrañaga, P.; De la Fuente, E.; Suarez, R.

2016-08-01

Extensive SEM work was carried out on deep-etched specimens to reveal the evolution of compacted and chunky graphite in magnesium-modified multicomponent Fe-C-Si alloys during early solidification and at room temperature. The findings of this research were then integrated in the current body of knowledge to produce an understanding of the crystallization of compacted and chunky graphite. It was confirmed that growth from the liquid for both compacted and chunky graphite occurs radially from a nucleus, as foliated crystals and dendrites. The basic building blocks of the graphite aggregates are hexagonal faceted graphite platelets with nanometer height and micrometer width. Thickening of the platelets occurs through growth of additional graphene layers nucleated at the ledges of the graphite prism. Additional thickening resulting in complete joining of the platelets may occur from the recrystallization of the amorphous carbon that has diffused from the liquid through the austenite, once the graphite aggregate is enveloped in austenite. With increasing magnesium levels, the foliated graphite platelets progressively aggregate along the c-axis forming clusters. The clusters that have random orientation, eventually produce blocky graphite, as the spaces between the parallel platelets disappear. This is typical for compacted graphite irons and tadpole graphite. The chunky graphite aggregates investigated are conical sectors of graphite platelets stacked along the c-axis. The foliated dendrites that originally develop radially from a common nucleus may aggregate along the c-axis forming blocky graphite that sometimes exhibits helical growth. The large number of defects (cavities) observed in all graphite aggregates supports the mechanism of graphite growth as foliated crystals and dendrites.

A graphite foam reinforced by graphite particles

Energy Technology Data Exchange (ETDEWEB)

Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

2007-11-15

Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

Graphite reactor physics; Physique des piles a graphite

Energy Technology Data Exchange (ETDEWEB)

Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm{sup 2}, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la

Graphite

Science.gov (United States)

Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

Redistribution and Effect of Various Elements on the Morphology of Primary Graphite in Cast Iron

Directory of Open Access Journals (Sweden)

J. Lacaze

2013-01-01

Full Text Available It has been shown repeatedly that many elements present as traces or at low level can affect graphite shape in cast irons. As part of a long term project aimed at clarifying the growth and the alteration of spheroidal graphite, a study on the effect of a few elements (Cu, Sn, Sb, and Ti on primary graphite growth was undertaken and analysed with reference to an alloy without any such additions. This work was performed by remelting alloys in graphite crucibles thus saturating the melt in carbon and enabling primary graphite to grow by controlled cooling of the melt above the eutectic temperature. Primary graphite growth in the reference alloy was observed to be lamellar, while the added elements were found to affect bulk graphite and to modify its outer shape, with Sb leading eventually to rounded agglomerates together with wavy lamellae. Secondary ion mass spectrometry was used to analyze the distribution of elements, and no build-up of trace elements at the graphite surface could be observed. Instead, it is established that the perturbation of bulk graphite is associated with inhomogeneous distribution of metallic elements inside graphite precipitates.

On estimating the fracture probability of nuclear graphite components

International Nuclear Information System (INIS)

Srinivasan, Makuteswara

2008-01-01

The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation

Leveraging comprehensive baseline datasets to quantify property variability in nuclear-grade graphites

Energy Technology Data Exchange (ETDEWEB)

Carroll, Mark C., E-mail: mark.carroll@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-2213 (United States); Windes, William E.; Rohrbaugh, David T. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-2213 (United States); Strizak, Joseph P.; Burchell, Timothy D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6088 (United States)

2016-10-15

Highlights: • An effort is underway to fully quantify the properties of nuclear-grade graphites. • Physical and mechanical properties of graphite are best characterized by distributions. • The Weibull distribution is most representative of graphite based on goodness-of-fit. • Fine-grained isomolded grades exhibit higher Weibull modulus values, indicative of more homogeneous properties. - Abstract: The full characterization of the physical and mechanical properties of candidate nuclear-grade graphites is highly dependent upon an understanding of the distribution of values that are inherent to graphite. Not only do the material properties of graphites vary considerably between grades owing to the raw materials sources, filler particle type and size, methods of compaction, and production process parameters, but variability is observed between billets of the same grade from a single batch and even across spatial positions within a single billet. Properly enveloping the expected properties of interest requires both a substantial amount of data to statistically capture this variability and a representative distribution capable of accurately describing the range of values. A two-parameter Weibull distribution is confirmed to be representative of the distribution of physical (density, modulus) and mechanical (compressive, flexure, and tensile strength) values in five different nuclear-grades of graphite. The fine-grained isomolded grades tend toward higher Weibull modulus and characteristic strength values, while the extruded grade being examined exhibits relatively large distributions in property values. With the number of candidate graphite specimens that can undergo full irradiation exposure and subsequent testing having limited feasibility with regard to economics and timely evaluations, a proper capture of the raw material variability in an unirradiated state can provide crucial supplementary resolution to the limited amount of available data on irradiated

Growth in children following irradiation for bone marrow transplantation

International Nuclear Information System (INIS)

Bushhouse, S.; Ramsay, N.K.; Pescovitz, O.H.; Kim, T.; Robison, L.L.

1989-01-01

Longitudinal height data from 46 pediatric bone marrow transplant (BMT) patients, including 18 with aplastic anemia (AA), 19 with acute nonlymphoblastic leukemia (ANLL), and 9 with acute lymphoblastic leukemia (ALL), were analyzed to assess growth posttransplantation. Patients were prepared for BMT with high-dose cyclophosphamide followed by 7.5 Gy single-dose irradiation; AA patients received total lymphoid irradiation (TLI), and leukemia patients received total body irradiation (TBI). AA patients demonstrated reduced height posttransplant as reflected in a negative mean standard deviation score. The observed reduction was statistically significant only at 3 years following transplant. In contrast, leukemia patients showed a significant loss in relative height that was first visible at 1 year post-BMT and continued until at least 4 years post-BMT. Mean growth velocities in the leukemia patients were significantly below median for the 3 years following transplant. With a median follow-up of 4 years, antithymocyte globulin plus steroids in combination with methotrexate as graft vs. host prophylaxis was associated with less severe growth suppression than methotrexate alone, while there were no significant associations between growth during the first 2 years following transplant and prepubertal status at transplant (as defined by age), graft vs. host disease, thyroid or gonadal function, or previous therapies received by the leukemia patients. Children undergoing marrow transplantation, particularly those receiving TBI, are at significant risk of subsequent growth suppression

Graphite tail powder and liquid biofertilizer as trace elements source for ground nut

Science.gov (United States)

Hindersah, Reginawanti; Setiawati, M. Rochimi; Fitriatin, B. Natalie; Suryatama, Pujawati; Asmiran, Priyanka; Panatarani, Camellia; Joni, I. Made

2018-02-01

Utilization of graphite tail waste from the mineral beneficiation processing is very important since it contain significant amount of essential minerals which are necessary for plant growth. These mineral are required in biochemical processes and mainly play an important role as cofactor in enzymatic reaction. The objective of this research is to investigate the performance of graphite tail on supporting plant growth and yield of ground nut (Arachishypogeae L.). A field experiment has been performed to test the performance of mixed graphite tail and reduced organic matter dose. The graphite tail size were reduced to various sieved size, -80 mesh, -100 mesh and -200 mesh. The experiment was setup in randomized block design with 4 treatments and 6 replications for each treatment, while the control plot is received without graphite tail. The results demonstrated that reduced organic matter along with -200 mesh tail has potentially decreased plant height at the end of vegetative growth stage, in contrast for to -80 mesh tail amendment increased individual fresh plant biomass. Statistically, there was no change of plant nodule, individual shoot fresh and dry weight, root nodule, number of pod following any mesh of graphite tail amendment. Reducing organic matter while adding graphite tail of 5% did not change bean weight in all plot. In contrast, reduced organic matter along with 80-mesh graphite tail amendment improved the nut yield per plot. This experiment suggests that graphite tail, mainly -80 mesh graphite tail can be possibly used in legume production.

UV-B-irradiation effect on growth reactions of phytopathogenic fungus fusarium solani

International Nuclear Information System (INIS)

Gushcha, M.Yi.; Dyachenko, A.Yi.; Dmitryijev, O.P.

2002-01-01

The UV-B irradiation effect on spore germination and hyphae growth of phythopathogenic fungus Fusarium solani was studied. Spores irradiation by small doses of 0,1 - 1,0 kJ/m 2 results in growth stimulation of primary hyphae. Adaptive effect of UV-B small doses for fungi was shown. Preliminary irradiation in doses of 0,1 - 0,5 kJ/m 2 increased spore radioresistance and diminished the effect of the next damaging dose

The effect of neutron irradiation on the trapping of tritium in carbon-based materials

International Nuclear Information System (INIS)

Kwast, H.; Werle, H.; Glugla, M.; Wu, C.H.; Federici, G.

1993-11-01

Carbon-based materials are considered for protection of plasma facing components in the next step fusion device. To investigate the effects of neutron damage on the tritium behaviour an experimental study on the tritium retention of various neutron irradiated graphites and carbon/carbon fibre composites was started. The irradiation dose of the specimens ranges from 10 -3 to 3.5 dpa.g and the irradiation temperature from 390 C to 1500 C. A comparison of tritium retention in pre- and post-irradiated carbon-based materials as a function of the sample temperature is reported in this paper and the results are discussed. The first results indicate that the retention of tritium is higher in irradiated graphite than in unirradiated graphite and depends largely on the density and microstructure. The retention is also influenced by the tritium-loading temperature. Graphite of type S 1611, irradiated at 400 C and 600 C up to a damage of 0.1 dpa.g, retained about two times more tritium than the unirradiated material. (orig.)

Nucleation/Growth Mechanisms and Morphological Evolution of Porous MnO2 Coating Deposited on Graphite for Supercapacitor

Directory of Open Access Journals (Sweden)

Wenxin Huang

2017-10-01

Full Text Available The nucleation and growth mechanisms of porous MnO2 coating deposited on graphite in MnSO4 solution were investigated in detail by cyclic voltammetry, chronoamperometry and scanning electron microscopy. The electrochemical properties of honeycomb-like MnO2 were evaluated by cycle voltammetry and galvanostatic charge-discharge. Results indicated that MnO2 was synthesized by the following steps: Mn2+→ Mn3++ e-, Mn3++2H2O → MnOOH + 3H+, and MnOOH → MnO2 + H++ e-. The deposition of MnO2 was divided into four stages. A short incubation period (approximately 1.5 s was observed, prior to nucleation. The decreasing trend of the current slowed as time increased due to nucleation and MnO2 growth in the second stage. A huge number of nuclei were formed by instantaneous nucleation, and these nuclei grew and connected with one another at an exceedingly short time (0.5 s. In the third stage, the gaps in-between initial graphite flakes were filled with MnO2 until the morphology of the flakes gradually became similar to that of the MnO2-deposited layer. In the fourth stage, the graphite electrode was covered completely with a thick and dense layer of MnO2 deposits. All MnO2 electrodes at different deposition times obtained nearly the same specific capacitance of approximately 186 F/g, thus indicating that the specific capacitance of the electrodes is not related with deposition time.

Stable Carbon Isotope Ratio (δ13C Measurement of Graphite Using EA-IRMS System

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Andrius Garbaras

2015-06-01

Full Text Available δ13C values in non-irradiated natural graphite were measured. The measurements were carried out using an elemental analyzer combined with stable isotope ratio mass spectrometer (EA-IRMS. The samples were prepared with ground and non-ground graphite, the part of which was mixed with Mg (ClO42. The best combustion of graphite in the oxidation furnace of the elemental analyzer was achieved when the amount of pulverized graphite ranged from 200 to 490 µg and the mass ratio C:Mg(ClO42 was approximately 1:10. The method for the graphite burning avoiding the isotope fractionation is proposed.DOI: http://dx.doi.org/10.5755/j01.ms.21.2.6873

Structural features of the adsorption layer of pentacene on the graphite surface and the PMMA/graphite hybrid surface

Science.gov (United States)

Fadeeva, A. I.; Gorbunov, V. A.; Litunenko, T. A.

2017-08-01

Using the molecular dynamics and the Monte Carlo methods, we have studied the structural features and growth mechanism of the pentacene film on graphite and polymethylmethacrylate /graphite surfaces. Monolayer capacity and molecular area, optimal angles between the pentacene molecules and graphite and PMMA/graphite surfaces as well as the characteristic angles between the neighboring pentacene molecules in the adsorption layer were estimated. It is shown that the orientation of the pentacene molecules in the film is determined by a number of factors, including the surface concentration of the molecules, relief of the surface, presence or absence of the polymer layer and its thickness. The pentacene molecules adsorbed on the graphite surface keep a horizontal position relative to the long axis at any surface coverage/thickness of the film. In the presence of the PMMA layer on the graphite, the increase of the number of pentacene molecules as well as the thickness of the PMMA layer induce the change of molecular orientation from predominantly horizontal to vertical one. The reason for such behavior is supposed to be the roughness of the PMMA surface.

Some equipment for graphite research in swimming pool reactors

International Nuclear Information System (INIS)

Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

1964-01-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr

Assessment of the radiological inventory of EDF's graphite waste through an assimilation method

International Nuclear Information System (INIS)

Poncet, B.

2014-01-01

The definitive disposal of graphite from the decommissioned UNGG reactors (Chinon A3, Saint-Laurent A1, Saint-Laurent A2 and Bugey 1) has required a radiological inventory of the irradiated graphite. This study focuses on Cl 36 that is produced by neutron absorption on Cl 35 that was present initially in graphite as an impurity (about 80 mg/t of Cl initially in Bugey 1 graphite)). It appears that the changes of Cl 36 concentration along the height of a stack of graphite do neither fit the changes in the neutron flux nor the changes in the graphite temperature. This fact is explained by the high level of purity of the graphite and the nugget effect. Challenged by the absence of spatial correlation of the Cl 36 concentration, an EDF's team has developed an assimilation method based on comparisons between calculations and measurements in order to get a conservative inventory. (A.C.)

Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

International Nuclear Information System (INIS)

Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

2014-01-01

Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)

irradiation growth in annealed Zr2.5wt%Nb at 3530K

International Nuclear Information System (INIS)

Rogerson, A.; Murgatroyd, R.A.

1978-10-01

Zr 2.5wt%Nb growth specimens have been irradiated at 353 0 K to a fast neutron dose of approximately 4.0 x 10 25 n/m 2 . Specimens were taken from the longitudinal and transverse directions of a nominally annealed, seam-welded tube and irradiated in both the stress relieved and fully annealed conditions. Growth in these specimens is characterised by large positive and negative strains in the longitudinal and transverse directions respectively, with dimensional changes in weld material exhibiting intermediate growth behaviour. The results are compared with growth data on both annealed and cold worked Zircaloy-2 at 353 0 K and discussed in terms of the effect of texture, grain size, and cold work on irradiation growth. It is concluded that the continuation of growth to high doses in annealed Zr-2.5wt%Nb at 353 0 K results from interstitial induced dislocation climb with vacancies diffusing to grain boundaries. (author)

Microstructural evolutions of zirconium alloys under irradiation. Link with the irradiation growth phenomenon

International Nuclear Information System (INIS)

Simonot, C.

1995-01-01

This study deals with the irradiation-induced growth and microstructural evolutions of Zircaloy-4 type materials (ZrSn 1.2-1.7 Fe 0.18-0.24 Cr 0.07-0.13 O 0.09-0.15 ), used as cladding and guide-tubes in PWR's fuel assemblies. The main objective was to obtain a better understanding of the growth acceleration which may occur at high doses for the recrystallized metallurgical state. The elongation values of stress-free tubes irradiated at 400 deg in experimental reactors give clear indication of accelerated growth after a critical dose. Microstructural investigations reveal some large basal plane dislocation loops with vacancy character, which is an unexpected defect configuration for an hexagonal material with a c/a ratio less than the ideal value. In addition, a significant redistribution of iron and chromium solute elements comes from the dissolution of the initial Zr(Fe,Cr) 2 phases. In a guide-tube irradiated to high dose at 320 deg in a power reactor, a large density of these c-component loops was also observed in coincidence with a large iron re-solution due to the progressive partial amorphization of Laves phases. By contrast, as long as a negligible amount of iron is available in the matrix (start of progressive amorphization at 350 deg or complete amorphization without any chemical change at 280 deg, only prism plane loops with interstitial and vacancy character are observed and the steady-state growth rate is low. A mechanism taking into account the Diffusional Anisotropy Difference of the radiation induced point defects seems to be the most suitable to explain the correlations between microstructural evolutions and growth rates. However it does not allow to predict the dose necessary for the formation of the basal plane loops responsible for the growth acceleration. Several factors (dose, temperature, metallurgical state) are required and the iron re-solution is likely to play a major role to stabilize such defects, by a decrease in stacking fault energy

Role of internal stresses in the transient of irradiation growth of zircaloy-2

International Nuclear Information System (INIS)

Tome, C.N.; Christodoulou, N.; Turner, P.A.; Miller, M.A.; Woo, C.H.; Root, J.; Holden, T.M.

1995-07-01

A 'self-consistent' polycrystalline model is used to simulate irradiation growth of Zircaloy-2 samples irradiated at about 330 K. The predictions of the model are compared with experimental measurements obtained from specimens irradiated in the Advanced Test Reactor (ATR) at Idaho Falls. Three types of material are studied here: annealed, cold worked in tension and cold worked by rolling. In general, the growth rate attains a steady-state value after it goes through a transient that depends on the initial state of the material. The transient growth behaviour is explained in terms of the evolution of intergranular residual stresses that are present in the sample, and in terms of the dislocation structure. From this study, information regarding irradiation creep and growth mechanisms occurring at the single crystal level is obtained. (author). 28 refs., 1 tab., 4 figs

Growth of cells superinoculated onto irradiated and nonirradiated confluent monolayers

International Nuclear Information System (INIS)

Matsuoka, H.; Ueo, H.; Sugimachi, K.

1990-01-01

We prepared confluent monolayers of normal BALB/c 3T3 cells and compared differences in the growth of four types of cells superinoculated onto these nonirradiated and irradiated monolayers. The test cells were normal BALB/c 3T3 A31 cells, a squamous cell carcinoma from a human esophageal cancer (KSE-1), human fetal fibroblasts, and V-79 cells from Chinese hamster lung fibroblasts. Cell growth was checked by counting the cell number, determining [3H]thymidine incorporation and assessing colony formation. We found that on nonirradiated monolayers, colony formation of human fetal fibroblasts and normal BALB/c 3T3 cells was completely inhibited. On irradiated cells, test cells did exhibit some growth. KSE-1 cells, which had a low clonogenic efficiency on plastic surfaces, formed colonies on both irradiated and nonirradiated cells. On these monolayers, the clonogenic efficiency of V-79 cells was also higher than that on plastic surfaces. We conclude that the nonirradiated monolayer of BALB/c 3T3 cells completely inhibits the growth of superinoculated normal BALB/c 3T3 and human fetal fibroblasts, while on the other hand, they facilitate the growth of neoplastic KSE-1 and V-79 cells by providing a surface for cell adherence and growth, without affecting the presence of normal cells in co-cultures

Effect of whole-body irradiation on skeletal growth in rhesus monkeys

International Nuclear Information System (INIS)

Sonneveld, P.; van Bekkum, D.W.

1979-01-01

Late effects of single whole-body doses of 400 to 500 and 750 to 900 rads on skeletal growth in 32 rhesus monkeys were studied. Findings indicated growth inhibition strongly related to dose and age at irradiation. Doses of 750 to 900 rads before the age of 40 months resulted in significantly greater growth inhibition (11%) than doses given during or shortly after adolescence (p < 0.005). Doses of less than 750 rads were not significant. In view of the close similarity between monkeys and man, irradiation of children at doses greater than 750 rads may carry a strong risk of subsequent growth retardation

Improved growth response to GH treatment in irradiated children

International Nuclear Information System (INIS)

Lannering, B.; Albertsson-Wikland, K.

1989-01-01

The growth response to two years of GH treatment was studied in fifteen children after radiotherapy for a cranial tumour. The growth response was compared to that of short children (-2 SD) and that of children with idiopathic growth hormone deficiency (GHD) of similar ages. All children were treated with hGH 0.1 IU/kg/day s.c.; which is a higher dose and frequency than previously reported for irradiated children. On this protocol the growth rate increased 5.0 +- 0.5 cm/y (mean +- SEM) the first year and 3.8 +- 0.7 cm/y the second year compared to the growth rate the year before GH-treatment. Although the net gain in growth was higher than previously reported, the first year growth response was significantly reduced (p less than 0.05) compared to that of GHD-children (7.6 +- 0.5 cm/y) but exceeded (p less than 0.05) that of short children (3.4 +- 0.3 cm/y). The median spontaneous 24 h-GH secretion was 209 mU/l in the short children, 52 mU/l in the irradiated children and 16 mU/l in the idiopathic GHD children. Thus the growth increment varied inversely to the spontaneous GH secretion observed in the three groups

In situ observations of graphite formation during solidification of cast iron

DEFF Research Database (Denmark)

Bjerre, Mathias Karsten

solidification and growth continues throughout solid state cooling and the eutectoid transformation. Years of research have greatly improved the understanding of the basic mechanisms that control graphite growth as well as the ability to control graphite morphology during industrial production of cast components......, the solidification of cast iron is studied with focus on formation and growth of spheroidal graphite. To this end, an experiment is conducted at the Diamond Light Source synchrotron facility in Harwell, UK: Employing an environmental cell devel-oped at the Manchester X-ray Imaging Facility at the University...... state growth presented in the present thesis. From the analysis it is clear that the presented data is of an unprecedented quality and that it represents a solid basis for validation of future models. Solidification simulations of a ductile cast iron component highlights the importance of the nucleation...

Response to growth hormone treatment and final height after cranial or craniospinal irradiation

International Nuclear Information System (INIS)

Sulmont, V.; Brauner, R.; Fontoura, M.; Rappaport, R.

1990-01-01

Growth hormone (GH) deficiency (GHD) induced by cranial irradiation has become a frequent indication of hGH substitutive therapy. This study analyses the growth response to hGH therapy and the factors involved in the decrease in growth velocity observed after cranial irradiation. One hundred children given cranial radiation for pathology distant from the hypothalamo-pituitary area were studied. Fifty-six of them received hGH therapy for GHD resulting in decreased growth velocity. The initial annual height gain in the cranial-irradiated group was comparable to that of patients treated for idiopathic GHD; additional spinal irradiation significantly reduced the growth response. Twenty-eight hGH-treated patients reached final heights which were compared to those of 2 untreated irradiated groups, one with GHD (n=27) and the other with normal GH secretion (n=17). The height SD score changes observed in hGH therapy were +0.3 in the cranial (n=10) and -1.2 SD in the craniospinal (n=18) groups. GH deficiency had contributed to a mean height loss of 1 SD and spinal irradiation to a loss of 1.4SD. The small effect of hGH therapy on final height is probably linked to the small bone age retardation at onset of hGH therapy and to the fact that irradiated children entered puberty at a younger age in terms of chronological age and bone age than the idiopathic GHD patients. These data suggest that the results of gGH therapy in irradiated children might be improved with higher and more fractionated hGH doses and, in some patients, by delaying puberty using luteinizing hormone releasing hormone analogs

Response to growth hormone treatment and final height after cranial or craniospinal irradiation

Energy Technology Data Exchange (ETDEWEB)

Sulmont, V.; Brauner, R.; Fontoura, M.; Rappaport, R. (Hospital des Enfants Malades, Paris (France). Pediatric Endocrinology Unit and INSERM U30)

1990-01-01

Growth hormone (GH) deficiency (GHD) induced by cranial irradiation has become a frequent indication of hGH substitutive therapy. This study analyses the growth response to hGH therapy and the factors involved in the decrease in growth velocity observed after cranial irradiation. One hundred children given cranial radiation for pathology distant from the hypothalamo-pituitary area were studied. Fifty-six of them received hGH therapy for GHD resulting in decreased growth velocity. The initial annual height gain in the cranial-irradiated group was comparable to that of patients treated for idiopathic GHD; additional spinal irradiation significantly reduced the growth response. Twenty-eight hGH-treated patients reached final heights which were compared to those of 2 untreated irradiated groups, one with GHD (n=27) and the other with normal GH secretion (n=17). The height SD score changes observed in hGH therapy were +0.3 in the cranial (n=10) and -1.2 SD in the craniospinal (n=18) groups. GH deficiency had contributed to a mean height loss of 1 SD and spinal irradiation to a loss of 1.4SD. The small effect of hGH therapy on final height is probably linked to the small bone age retardation at onset of hGH therapy and to the fact that irradiated children entered puberty at a younger age in terms of chronological age and bone age than the idiopathic GHD patients. These data suggest that the results of gGH therapy in irradiated children might be improved with higher and more fractionated hGH doses and, in some patients, by delaying puberty using luteinizing hormone releasing hormone analogs.

Development of a rotating graphite carbon disk stripper

Science.gov (United States)

Hasebe, Hiroo; Okuno, Hiroki; Tatami, Atsushi; Tachibana, Masamitsu; Murakami, Mutsuaki; Kuboki, Hironori; Imao, Hiroshi; Fukunishi, Nobuhisa; Kase, Masayuki; Kamigaito, Osamu

2018-05-01

Highly oriented graphite carbon sheets (GCSs) were successfully used as disk strippers. An irradiation test conducted in 2015 showed that GCS strippers have the longest lifetime and exhibit improved stripping and transmission efficiencies. The problem of disk deformation in previously used Be-disk was solved even with higher beam intensity.

Nonlinear seismic analysis of a graphite reactor core

International Nuclear Information System (INIS)

Laframboise, W.L.; Desmond, T.P.

1988-01-01

Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

The effects of irradiation and temperature on the growth of Zircaloy-4 tubes

International Nuclear Information System (INIS)

Kendoush, A.A.

1987-01-01

The growth strain was measured after irradiation for 16 Zircaloy-4 tubes of the recrystallised and stress relieved types. The operating temperature during irradiation ranged between 317 and 344 0 C. The average fast neutron fluence was 9.6x10 20 n/cm 2 . Experimental results indicated the dependence of the growth on the irradiation temperature. The stress relieved result was compared with data of the literature. (orig.)

Inhibitory Effect of Gamma-Irradiated Chitosan on the Growth of Denitrifiers

Directory of Open Access Journals (Sweden)

Javier Vilcáez

2009-01-01

Full Text Available In order to find an environmentally benign substitute to hazardous inhibitory agents, the inhibitory effect of -irradiated chitosans against a mixed culture of denitrifying bacteria was experimentally evaluated. Unlike other studies using pure aerobic cultures, the observed effect was not a complete inhibition but a transient inhibition reflected by prolonged lag phases and reduced growth rates. Raw chitosan under acid conditions (pH 6.3 exerted the strongest inhibition followed by the 100 kGy and 500 kGy irradiated chitosans, respectively. Therefore, because the molecular weight of chitosan decreases with the degree of -irradiation, the inhibitory properties of chitosan due to its high molecular weight were more relevant than the inhibitory properties gained due to the modification of the surface charge and/or chemical structure by -irradiation. High dosage of -irradiated appeared to increase the growth of mixed denitrifying bacteria in acid pH media. However, in neutral pH media, high dosage of -irradiation appeared to enhance the inhibitory effect of chitosan.

Growth in children treated for acute lymphoblastic leukemia with and without prophylactic cranial irradiation

International Nuclear Information System (INIS)

Moell, C.; Garwicz, S.; Marky, L.; Melander, L.; Karlberg, J.

1988-01-01

Growth and weight gain were studied longitudinally over a period of four years in thirty-nine children treated for acute lymphoblastic leukemia. The children were divided into two groups according to treatment. Twenty-eight children were given prophylactic cranial irradiation and eleven children were treated without such irradiation. The duration of cytostatic treatment was three years in all cases. Average growth during the first two years was similar in the two groups, and the standard deviation scores (SDS) were below average. The rate of growth during the fourth year was significantly higher among those children who had not received cranial irradiation. After four years the average attained height had declined 0.5 SD for children treated with cranial irradiation and 0.2 SD for children without such treatment. Attained weight after four years had increased 0.4 SD more among those children who had not received irradiation. The results suggest that prophylactic cranial irradiation is responsible for the greater part of the prepubertal growth inhibition in these children. (authors)

Experimental observation of the layering and wetting of multilayer liquid helium-4 films on graphite

International Nuclear Information System (INIS)

Kumar, S.

1987-01-01

The multilayer adsorption of liquid 4 He on graphite was studied by using third sound, a substrate-induced surface wave in a superfluid film, to probe the 4 He film-vapor interface. The third-sound velocity decreases with increasing film thickness and can be used to monitor the film growth. Graphite, forms of which have large areas of atomic uniformity, is an ideal substrate for the study of film growth and layering. An annular resonator made out of graphite fibers was used for the experiments. Such a resonator avoids problems such as capillary condensation present in earlier resonance experiments with graphite foam and vapor sound interference present in time-of-flight experiments with highly oriented pyrolitic graphite (HOPG). Measurements of film growth were made between temperatures of 0.35 and 1.25 K. The third-sound resonance frequency, which is proportional to the third-sound velocity, was used to follow the film growth. Simultaneous measurements of the third-sound velocity on glass provide an independent measure of the film thickness. Results obtained show continuous film growth up to at least 25 to 30 layers on graphite. Oscillations of the third-sound velocity for low film coverages shown evidence of layering of the film

Long Term Behaviour of 14C and Stability Assessments of Graphite Under Repository Conditions

International Nuclear Information System (INIS)

Jones, Abbie N.; McDermott, Lorraine; Worth, Robert; Hagos, Bereket; Black, Greg; Marsden, Barry J

2016-01-01

The key objectives of the University of Manchester’s nuclear graphite research within the CRP are to provide analysis on the long term behaviour and stability assessments of irradiated graphite waste. The research will concentrate on isotopic 14 C mobility under repository environments. This also requires an understanding the long-term behaviour of the final waste form under repository conditions. Procedures to evaluate the long term leaching properties of radionuclides from irradiated graphite waste has been developed by combining ANSI 16.1 (USA) and NEN 7345 (Netherlands) standardised diffusion leaching techniques. The ANSI 16.1 standard has been followed to acquire the leachates and to determine the leach rate and diffusion coefficient. The NEN 7345 standard technique has been used to determine the diffusion mechanism of radionuclides. The investigation employs simulated Drigg groundwater as a leachant using semi-dynamic technique for the production of leachate specimens. Analysis of 3 H and 14 C activity release from Magnox graphite was measured using liquid scintillating counting. Preliminary results show that there is an initial high release of activity and decreases when the leaching period increases. This may be due to the depletion of contaminants that were initially bound by the internal pore networks and the free surface. During the leaching test approximately 275.33 ± 18.20 Bq of 3 H and 106.26 ± 7.01 Bq of 14 C was released into the leachant within 91 days. The work reported herein contributed several key findings to the international work on graphite leaching to offer guidance leading toward obtaining leaching data in the future: (a) the effective diffusion coefficient for 14 C from graphite waste has been determined. The diffusion process for 14 C has two stages resulting two different values of diffusion coefficient, i.e., for the fast and slow components; (b) the controlling leaching mechanism for 3 H radionuclide from graphite is shown to be

A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

International Nuclear Information System (INIS)

Birch, M.; Brocklehurst, J.E.

1987-12-01

The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)

IAEA International Database on Irradiated Nuclear Graphite Properties. Summary report of consultants' meeting. 12. meeting of the Technical Steering Committee

International Nuclear Information System (INIS)

Chung, H.K.; Wickham, A.J.

2010-02-01

The 12th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 12-13 November 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

Effect of gamma-irradiation to the embryos growth of kenaf and roselle

International Nuclear Information System (INIS)

Matsuzaka, Teruki; Okuyama, Naoki; Fujino, Hideki

2008-01-01

Kenaf and roselle are widely cultivated in many countries and it has been used to produce several products or beverages. Also, these plants are recognized as an environmental friendly crop. The gamma-ray irradiation was commonly used to avoid the spoilage of several vegetables or to inhibit the emergence of larvae from seeds. Recently, radiation breeding was characterized for the creation of new varieties and the improvement of vegetative propagation. As of today, little studies have been reported regarding the radiation sensitivity of these plants. To gain a better understanding on the embryos growth of kenaf and roselle, gamma-ray irradiation was performed on these seeds using 60Co irradiator. A statistically significance was observed on germination and embryo growth at irradiated kenaf seeds between control groups. In case of roselle, there was not statistically significant difference in the germination of irradiated seeds between control groups at the same doses. Moreover, the several truncated DNA forms were detected at the irradiated seeds and their embryos in electrophoresis assay. Based on these results, inhibitory effect on germination and embryo growth were closely related to the DNA damages. The present study demonstrates that remarkable difference on the radiation tolerance of seed was found between kenaf and roselle. (author)

Effect of low level prenatal X-irradiation on postnatal growth in the Wistar rat

International Nuclear Information System (INIS)

Jensh, R.P.; Brent, R.L.

1988-01-01

Forty-five pregnant Wistar strain rats were exposed to 0.0, 0.4, 0.6, or 0.8 Gy X-radiation on the 9th or 17th day of gestation to determined if prenatal X-irradiation would result in alterations in postnatal growth or growth rate. The mothers delivered their offspring, and the litters were reduced to a maximum of eight per litter on the second postnatal day. The 336 offspring were weighed weekly from day 3 until day 86, at which time they were killed, an autopsy was performed, and selected organs were removed and weighed. Postnatal growth rates did not differ significantly in irradiated offspring compared to sham irradiated animals. Irradiation on the 9th day, at any of the 3 dosage levels, did not result in significant differences in weekly weight. Weekly weight remained significantly lower due to irradiation on the 17th day of gestation. The gonadal weight ratio was significantly reduced in males irradiated on the 9th day. There were not other statistically significant changes in organ weight or organ/body weight ratios due to these levels of prenatal X-irradiation on the 9th or 17th day of pregnancy. These results indicate that low level prenatal X-irradiation, on the 17th day of rat gestation, causes prenatal growth retardation, evident at birth, which is not recuperable postnatally. Exposure to x-radiation at this time, however, does not affect the rate at which offspring grow during postnatal life. Offspring are smaller because they never fully recover from the original radiation-induced prenatal growth retardation

Inhomogeneous microstructural growth by irradiation

DEFF Research Database (Denmark)

Krishan, K.; Singh, Bachu Narain; Leffers, Torben

1985-01-01

In the present paper we discuss the development of heterogeneous microstructure for uniform irradiation conditions. It is shown that microstructural inhomogeneities on a scale of 0.1 μm can develop purely from kinematic considerations because of the basic structure of the rate equations used...... to describe such evolution. Two aspects of the growth of such inhomogeneities are discussed. Firstly, it is shown that a local variation in the sink densities of the various microstructural defects will tend to enhance the inhomogeneity rather than remove it. Secondly, such inhomogeneities will lead to point...... defect fluxes that result in a spatial growth of the inhomogeneous region, which will be stopped only when the microstructural density around this region becomes large. The results have important implications in the formation of denuded zones and void formation in metals....

An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

1991-05-01

The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

The impact of cranial irradiation on the growth of children with acute lymphocytic leukemia

International Nuclear Information System (INIS)

Wells, R.J.; Foster, M.B.; D'Ercole, A.J.; McMillan, C.W.

1983-01-01

Heights, height velocities, weights, and weight velocities were measured serially in 21 patients with acute lymphocytic leukemia (ALL) who had survived three to five years in continuous complete remission. These patients were assigned randomly to treatment regimens that varied according to whether cranial irradiation was used. Patients receiving cranial irradiation had lower height velocities during therapy than normal subjects and patients not receiving cranial irradiation. Twenty-two other children with ALL, who were irradiated but not randomized, exhibited similar alterations in growth. These results indicate that cranial irradiation, and not leukemia or antileukemia chemotherapy, causes reduced growth

Rheological and electrical properties of hybrid nanocomposites of epoxy resins filled with graphite nanoplatelets and carbon black.

Science.gov (United States)

Truong, Quang-Trung; Lee, Seon-Suk; Lee, Dai-Soo

2011-02-01

Graphite nanoplatelets (GNP) were prepared by microwave irradiation of natural graphites intercalated with ferric chloride in nitromethane (GIC). Intercalated structure of GIC was confirmed by X-ray diffraction patterns. SEM images of GIC after microwave irradiation showed the exfoliation of GIC, the formation of GNPs. Hybrid nanocomposites of bisphenol-A type epoxy resins filled with GNP and a conductive carbon black (CB) were prepared and rheological and electrical properties of the nanocomposites were investigated. Viscosity and electrical surface resistivity of the nanocomposites showed minima at certain mixtures of GNP and CB in the epoxy resins.

Hydrogen adsorption on and solubility in graphites

International Nuclear Information System (INIS)

Kanashenko, S.L.; Wampler, W.R.

1996-01-01

The experimental data on adsorption and solubility of hydrogen isotopes in graphite over a wide range of temperatures and pressures are reviewed. Langmuir adsorption isotherms are proposed for the hydrogen-graphite interaction. The entropy and enthalpy of adsorption are estimated, allowing for effects of relaxation of dangling sp 2 bonds. Three kinds of traps are proposed: edge carbon atoms of interstitial loops with an adsorption enthalpy relative to H 2 gas of -4.4 eV/H 2 (unrelaxed, Trap 1), edge carbon atoms at grain surfaces with an adsorption enthalpy of -2.3 eV/H 2 (relaxed, Trap 2), and basal plane adsorption sites with an enthalpy of +2.43 eV/H 2 (Trap 3). The adsorption capacity of different types of graphite depends on the concentration of traps which depends on the crystalline microstructure of the material. The number of potential sites for the 'true solubility' (Trap 3) is assumed to be about one site per carbon atom in all types of graphite, but the endothermic character of this solubility leads to a negligible H inventory compared to the concentration of hydrogen in type 1 and type 2 traps for temperatures and gas pressures used in the experiments. Irradiation with neutrons or carbon atoms increases the concentration of type 1 and type 2 traps from about 20 and 200 appm respectively for unirradiated (POCO AXF-5Q) graphite to about 1500 and 5000 appm, respectively, at damage levels above 1 dpa. (orig.)

The relationship between irradiation induced dimensional change and the coefficient of thermal expansion: a modified Simmons relationship

International Nuclear Information System (INIS)

Hall, G.; Marsden, B.J.; Fok, S.L.; Smart, J.

2003-01-01

In the 1960s, a theoretical relationship between the dimensional changes and the coefficient of thermal expansion of irradiated graphite was derived by J.H.W. Simmons. The theory was shown to be comparable with experimental observations at low irradiation doses, but shown to diverge at higher irradiation doses. However, various modified versions of this theory have been used as the foundation of design and life prediction calculations for graphite-moderated reactors. This paper re-examines the Simmons relationship, summarising its derivation and assumptions. The relationship was then modified to incorporate the high dose, high strain changes that were assumed to be represented in the changes in Young's modulus with irradiation dose. By scrutinising the behaviour of finite element analyses, it was possible to use a modified Simmons relationship to predict the dimensional changes of an isotropic and anisotropic graphite to high irradiation doses. These issues are important to present high-temperature reactors (HTRs) as the life of HTR graphite components is dependent upon their dimensional change behaviour. A greater understanding of this behaviour will help in the selection and development of graphite materials

Nano-cracks in a synthetic graphite composite for nuclear applications

Science.gov (United States)

Liu, Dong; Cherns, David

2018-05-01

Mrozowski nano-cracks in nuclear graphite were studied by transmission electron microscopy and selected area diffraction. The material consisted of single crystal platelets typically 1-2 nm thick and stacked with large relative rotations around the c-axis; individual platelets had both hexagonal and cubic stacking order. The lattice spacing of the (0002) planes was about 3% larger at the platelet boundaries which were the source of a high fraction of the nano-cracks. Tilting experiments demonstrated that these cracks were empty, and not, as often suggested, filled by amorphous material. In addition to conventional Mrozowski cracks, a new type of nano-crack is reported, which originates from the termination of a graphite platelet due to crystallographic requirements. Both types are crucial to understanding the evolution of macro-scale graphite properties with neutron irradiation.

Molecular dynamics study on interfacial thermal conductance of unirradiated and irradiated SiC/C

International Nuclear Information System (INIS)

Wang, Qingyu; Wang, Chenglong; Zhang, Yue; Li, Taosheng

2014-01-01

SiC f /SiC composite materials have been considered as candidate structural materials for several types of advanced nuclear reactors. Both experimental and computer simulations studies have revealed the degradation of thermal conductivity for this material after irradiation. The objective of this study is to investigate the effect of SiC/graphite interface structure and irradiation on the interfacial thermal conductance by using molecular dynamics simulation. Five SiC/graphite composite models were created with different interface structures, and irradiation was introduced near the interfaces. Thermal conductance was calculated by means of reverse-NEMD method. Results show that there is a positive correlation between the interfacial energy and interfacial C–Si bond quantity, and irradiated models showed higher interfacial energy compared with their unirradiated counterparts. Except the model with graphite atom plane parallel to the interface, the interfacial thermal conductance of unirradiated and irradiated (1000 eV) models, increases as the increase of interfacial energy, respectively. For all irradiated models, lattice defects are of importance in impacting the interfacial thermal conductance depending on the interface structure. For the model with graphite layer parallel to the interface, the interfacial thermal conductance increased after irradiation, for the other models the interfacial thermal conductance decreased. The vibrational density of states of atoms in the interfacial region was calculated to analyze the phonon mismatch at the interface

Graphitic Carbon-Based Nanostructures for Energy and Environmental Applications

Science.gov (United States)

Chan, Ka Long Donald

This thesis focuses on the synthesis and characterization of graphitic carbonbased photocatalytic nanostructures for energy and environmental applications. The preparation of carbon- and oxygen-rich graphitic carbon nitride with enhanced photocatalytic hydrogen evolution property was investigated. Composite materials based on graphene quantum dots were also prepared. These composites were used for photocatalytic degradation of organic pollutants and photoelectrocatalytic disinfection. The first part of this thesis describes a facile method for the preparation of carbon- and oxygen-rich graphitic carbon nitride by thermal condensation. Incorporation of carbon and oxygen enhanced the photoresponse of carbon nitride in the visible-light region. After exfoliation, the product was c.a. 45 times more active than bulk graphitic carbon nitride in photocatalytic hydrogen evolution under visible-light irradiation. In the second part, a simple approach to enhance the photocatalytic activity of red phosphorus was developed. Mechanical ball milling was applied to reduce the size of red phosphorus and to deposit graphene quantum dots (GQDs) onto red phosphorus. The product exhibited high visible-light-driven photocatalytic performance in the photodegradation of Rhodamine B. The incorporation of GQDs in titanium dioxide could also extend the absorption spectrum of TiO2 into the visible-light range. The third part of this thesis reports on the fabrication of a visible-light-driven composite photocatalyst of TiO2 nanotube arrays (TNAs) and GQDs. Carboxyl-containing GQDs were covalently coupled to amine-modified TNAs. The product exhibited enhanced photocurrent and high photoelectrocatalytic performance in the inactivation of E. coli under visible-light irradiation. The role of various reactive species in the photoelectrocatalytic process was investigated.

Exfoliated graphite/titanium dioxide nanocomposites for photodegradation of eosin yellow

International Nuclear Information System (INIS)

Ndlovu, Thabile; Kuvarega, Alex T.; Arotiba, Omotayo A.; Sampath, Srinivasan; Krause, Rui W.; Mamba, Bhekie B.

2014-01-01

Graphical abstract: - Highlights: • Preparation of exfoliated graphite (EG) from natural graphite. • Sol–gel anchoring of TiO 2 on exfoliated graphite. • High adsorption and photoactivity was observed for the EG-TiO 2 nanocomposite. • Mechanism of enhancement was proposed. - Abstract: An improved photocatalyst consisting of a nanocomposite of exfoliated graphite and titanium dioxide (EG-TiO 2 ) was prepared. SEM and TEM micrographs showed that the spherical TiO 2 nanoparticles were evenly distributed on the surface of the EG sheets. A four times photocatalytic enhancement was observed for this floating nanocomposite compared to TiO 2 and EG alone for the degradation of eosin yellow. For all the materials, the reactions followed first order kinetics where for EG-TiO 2 , the rate constant was much higher than for EG and TiO 2 under visible light irradiation. The enhanced photocatalytic activity of EG-TiO 2 was ascribed to the capability of graphitic layers to accept and transport electrons from the excited TiO 2 , promoting charge separation. This indicates that carbon, a cheap and abundant material, can be a good candidate as an electron attracting reservoir for photocatalytic organic pollutant degradation

Helium generation and diffusion in graphite and some carbides

International Nuclear Information System (INIS)

Holt, J.B.; Guinan, M.W.; Hosmer, D.W.; Condit, R.H.; Borg, R.J.

1976-01-01

The cross section for the generation of helium in neutron irradiated carbon was found to be 654 mb at 14.4 MeV and 744 mb at 14.9 MeV. Extrapolating to 14.1 MeV (the fusion reactor spectrum) gives 615 mb. The diffusion of helium in dense polycrystalline graphite and in pyrographite was measured and found to be D = 7.2 x 10 -7 m 2 s -1 exp (-80 kJ/RT). It is assumed that diffusion is primarily in the basal plane direction in crystals of the graphite. In polycrystalline graphite the path length is a factor of √2 longer than the measured distance due to the random orientation mismatch between successive grains. Isochronal anneals (measured helium release as the specimen is steadily heated) were run and maximum release rates were found at 200 0 C in polycrystalline graphite, 1000 0 C in pyrographite, 1350 0 C in boron carbide, and 1350 0 and 2400 0 C (two peaks) in silicon carbide. It is concluded that in these candidates for curtain materials in fusion reactors the helium releases can probably occur without bubble formation in graphites, may occur in boron carbide, but will probably cause bubble formation in silicon carbide. 7 figures

Growth rate of dislocation loop in Fe-Ni-Cr alloy under Kr+ ion and electron irradiation

International Nuclear Information System (INIS)

Kimoto, T.; Allen, C.W.; Rehn, L.E.

1991-10-01

In order to examine the effect of irradiating particle species on the growth rate of radiation-induced dislocation loops, a solution-annealed Fe-25Ni-15Cr-0.02C alloy was irradiated at 723 K first by 1.5 MeV Kr + ions for 2520 sec, then by 1.5 MeV Kr + ions and 1.0 MeV electrons simultaneously for 780 sec, and finally by 1.0 MeV electrons for 780 sec with the HVEM-Tandem Facility in Argonne National Laboratory. The calculated damage rate by 1.5 MeV Kr + ions was 5.8 x 10 -4 dpa/s, and that by 1.0 MeV electrons was 1 x 10 -4 dpa/s. The growth rate of a dislocation loop located at the center of the specimen was 7 x 10 -3 nm/s for the Kr + ion irradiation, 4 x 10 -2 nm/s for the simultaneous Kr + and electron irradiation, and (2--3) x 10 -2 nm/s for the electron irradiation. This implies that the electron irradiation is about 19 times more effective in the growth of radiation-induced dislocation loops than the Kr + ion irradiation. The dislocation loop growth rate under the simultaneous Kr + and electron irradiation is higher than the sum of the growth rates under the individual Kr + and electron irradiations. 5 refs., 4 figs

Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

International Nuclear Information System (INIS)

Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

1977-10-01

Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

Energy Technology Data Exchange (ETDEWEB)

Davenport, Michael; Petti, D. A.

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

Effects of ultraviolet and electron radiations on graphite-reinforced polysulfone and epoxy resins

International Nuclear Information System (INIS)

Giori, C.; Yamauchi, T.

1984-01-01

Degradation mechanisms have been investigated for graphite/polysulfone and graphite/epoxy laminates exposed to ultraviolet and high-energy electron radiations in vacuum up to 960 equivalent sun hours and 10 9 rads, respectively. Based on GC and combined GC/MS analysis of volatile by-products evolved during irradiation, several free radical mechanisms of composite degradation have been identified. All the composite materials evaluated have shown high electron radiation stability and relatively low ultraviolet stability as indicated by low G values and high quantum yields for gas formation. Mechanical property measurements of irradiated samples did not reveal significant changes, with the possible exception of UV exposed polysulfone laminates. Hydrogen and methane have been identified as the main byproducts of irradiation, along with unexpectedly high levels of CO and CO 2 . Initial G values for methane relative to hydrogen formation are higher in the presence of isopropylidene linkages, which occur in bisphenol-A resins

Summary report of consultants meeting on IAEA International Database on Irradiated Nuclear Graphite Properties. 11. meeting of the Technical Steering Committee

International Nuclear Information System (INIS)

Humbert, D.; Wickham, A.J.

2009-05-01

The 11th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 25-26 March 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

Summary report of consultants' meeting - IAEA International Database on Irradiated Nuclear Graphite Properties. 8th meeting of the Technical Steering Committee

International Nuclear Information System (INIS)

Humbert, D.; Wickham, A.J.

2006-05-01

The '8th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 15-16 March 2006 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

Method for producing dustless graphite spheres from waste graphite fines

Science.gov (United States)

Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

2012-05-08

A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

Structural disorder of graphite and implications for graphite thermometry

Science.gov (United States)

Kirilova, Martina; Toy, Virginia; Rooney, Jeremy S.; Giorgetti, Carolina; Gordon, Keith C.; Collettini, Cristiano; Takeshita, Toru

2018-02-01

Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25 megapascal (MPa) and aseismic velocities of 1, 10 and 100 µm s-1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

OpenAIRE

Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

2012-01-01

Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

Effect of Graphite Doped TiO_2 Nanoparticles on Smoke Degradation

International Nuclear Information System (INIS)

Roshasnorlyza Hazan; Mohamad Shahrizal Md Zain; Natrah Syafiqah Rosli

2016-01-01

Secondhand smoke affects in the same way as regular smoker. The best solution is to purify the air efficiently and effectively. In this study, we were successfully doped TiO_2 nanoparticle with graphite to accelerate the degradation of cigarette smoke. The graphite doped and undoped TiO_2 nanoparticles were prepared from synthetic rutile using alkaline fusion method and their photo catalytic activity were investigated under visible light irradiation. The photo catalytic activity of the TiO_2 nanoparticles was analyzed in terms of their particle size analysis, crystallization and optical band gap. TiO_2 nanoparticle act as photo catalyzer by utilization of light energy to excite electron-hole pairs in smoke degradation processes. With the aided from graphite in TiO_2 nanoparticles, the smoke degradation was accelerate up to 44.4 %. In this case, graphite helps to reduce optical band gap of TiO_2 nanoparticle, thus increasing excitation of electron from valence band to conduction band. (author)

Germanium Nanowires-in-Graphite Tubes via Self-Catalyzed Synergetic Confined Growth and Shell-Splitting Enhanced Li-Storage Performance.

Science.gov (United States)

Sun, Yong; Jin, Shuaixing; Yang, Guowei; Wang, Jing; Wang, Chengxin

2015-04-28

Despite the high theoretical capacity, pure Ge has various difficulties such as significant volume expansion and electron and Li(+) transfer problems, when applied as anode materials in lithium ion battery (LIB), for which the solution would finally rely on rational design like advanced structures and available hybrid. Here in this work, we report a one-step synthesis of Ge nanowires-in-graphite tubes (GNIGTs) with the liquid Ge/C synergetic confined growth method. The structure exhibits impressing LIB behavior in terms of both cyclic stability and rate performance. We found the semiclosed graphite shell with thickness of ∼50 layers experience an interesting splitting process that was driven by electrolyte diffusion, which occurs before the Ge-Li alloying plateau begins. Two types of different splitting mechanism addressed as "inside-out"/zipper effect and "outside-in" dominate this process, which are resulted from the SEI layer growing longitudinally along the Ge-graphite interface and the lateral diffusion of Li(+) across the shell, respectively. The former mechanism is the predominant way driving the initial shell to split, which behaves like a zipper with SEI layer as invisible puller. After repeated Li(+) insertion/exaction, the GNIGTs configuration is finally reconstructed by forming Ge nanowires-thin graphite strip hybrid, both of which are in close contact, resulting in enormous enchantment to the electrons/Li(+) transport. These features make the structures perform well as anode material in LIB. We believe both the progress in 1D assembly and the structure evolution of this Ge-C composite would contribute to the design of advanced LIB anode materials.

Microarray Cluster Analysis of Irradiated Growth Plate Zones Following Laser Microdissection

International Nuclear Information System (INIS)

Damron, Timothy A.; Zhang Mingliang; Pritchard, Meredith R.; Middleton, Frank A.; Horton, Jason A.; Margulies, Bryan M.; Strauss, Judith A.; Farnum, Cornelia E.; Spadaro, Joseph A.

2009-01-01

Purpose: Genes and pathways involved in early growth plate chondrocyte recovery after fractionated irradiation were sought as potential targets for selective radiorecovery modulation. Materials and Methods: Three groups of six 5-week male Sprague-Dawley rats underwent fractionated irradiation to the right tibiae over 5 days, totaling 17.5 Gy, and then were killed at 7, 11, and 16 days after the first radiotherapy fraction. The growth plates were collected from the proximal tibiae bilaterally and subsequently underwent laser microdissection to separate reserve, perichondral, proliferative, and hypertrophic zones. Differential gene expression was analyzed between irradiated right and nonirradiated left tibia using RAE230 2.0 GeneChip microarray, compared between zones and time points and subjected to functional pathway cluster analysis with real-time polymerase chain reaction to confirm selected results. Results: Each zone had a number of pathways showing enrichment after the pattern of hypothesized importance to growth plate recovery, yet few met the strictest criteria. The proliferative and hypertrophic zones showed both the greatest number of genes with a 10-fold right/left change at 7 days after initiation of irradiation and enrichment of the most functional pathways involved in bone, cartilage, matrix, or skeletal development. Six genes confirmed by real-time polymerase chain reaction to have early upregulation included insulin-like growth factor 2, procollagen type I alpha 2, matrix metallopeptidase 9, parathyroid hormone receptor 1, fibromodulin, and aggrecan 1. Conclusions: Nine overlapping pathways in the proliferative and hypertrophic zones (skeletal development, ossification, bone remodeling, cartilage development, extracellular matrix structural constituent, proteinaceous extracellular matrix, collagen, extracellular matrix, and extracellular matrix part) may play key roles in early growth plate radiorecovery.

Assessments of the stresses and deformations in an RBMK graphite moderator brick

International Nuclear Information System (INIS)

Jones, C.J.; Davies, M.A.; Marsden, B.J.; Bougaenko, S.E.; Baldin, V.D.; Demintievski, V.N.; Rodtchenkov, B.S.; Sinitsyn, E.N.

1996-01-01

The RBMK reactors, designed by RDIPE (Moscow), are graphite moderated and cooled by light water. Graphite dimensions and thermo-mechanical properties change significantly in a complex manner during reactor life due to fast neutron damage and these changes have implications on the safe operation of all graphite moderated reactors. A joint programme of work is being carried out between AEA Technology (UK) and RDIPE (Russia) to assess the life of the RBMK graphite stack under normal operating conditions. The programme has included the modelling of graphite dimensional changes due to irradiation through reactor life and the assessment of the implications of these changes on the stresses and deformations in the graphite stack. Calculations have been carried out to assess the deformations of a moderator brick over a period from start of life up to 30 years of operation. The assessment have also included an analysis of the stresses in the bricks so that the time to brick failure could be determined. This paper describes the RBMK core design, the data and assessment methodology used in the analysis of the RBMK core and presents some results from analyses of the Leningrad Unit 1 RBMK reactor. (author). 2 refs, 8 figs

Host influence on irradiation bioefficacy : growth and development of Spodoptera litura (Fabricius)

International Nuclear Information System (INIS)

Seth, R.K.

1995-01-01

Survival and development of Spodoptera litura(F), a polyphagous pest was ascertained on a range of host plants, common and economic in the Indian context. Castor, cotton, green gram and okra constituted the preferred host group showing significantly better growth rate (GR) and growth index (GI) than the group of less preferred host plants that included groundnut, red gram, rose leaf and rose petal. Ontogenic growth profile vis-a-vis irradiation was studied with recording of GI at L3, pupa and adult levels in response to gamma dosages administered in L1 stage. Irradiation effect on the growth index was mainly due to reduction in survival, which was further pronounced owing to delay in development. On all the preferred host plants, adult-G1 was reduced by more than 50 per cent at 40 Gy gamma dose; whereas the same dose on less preferred host plants could prevent adult emergence on ground nut and red gram, and inhibited pupa formation on rose diet. The percentage mortality of larvae exhibiting delayed development was markedly more in the less preferred host group. Dose dependent reduction was observed in the larval growth rate of L3, L5 and L6 instars on all the host plants. 20 Gy exhibited almost same intensity of adverse impact in less preferred host group as was shown by 40 Gy in preferred hosts. 20 Gy dose reduced the pupal weight and conversion ratio of treated insects significantly. Age influence was evident on irradiation efficacy on all the diets. Insects treated in the later instar had less radiosusceptibility as compared to the young irradiated larvae. Irradiation had a negative correlation with survival and weight gain during the course of development that in turn bore a positive relation with the increase in radiation dosage. Food is an important governing factor in influencing the insects survival value and developmental behaviour and may modify its intrinsic sensitivity towards irradiation stress. (author). 21 refs., 1 fig., 5 tabs

Non-activated high surface area expanded graphite oxide for supercapacitors

Science.gov (United States)

Vermisoglou, E. C.; Giannakopoulou, T.; Romanos, G. E.; Boukos, N.; Giannouri, M.; Lei, C.; Lekakou, C.; Trapalis, C.

2015-12-01

Microwave irradiation of graphite oxide constitutes a facile route toward production of reduced graphene oxide, since during this treatment both exfoliation and reduction of graphite oxide occurs. In this work, the effect of pristine graphite (type, size of flakes), pretreatment and oxidation cycles on the finally produced expanded material was examined. All the types of graphite that were tested afforded materials with high BET surface areas ranging from 940 m2/g to 2490 m2/g, without intervening an activation stage at elevated temperature. SEM and TEM images displayed exfoliated structures, where the flakes were significantly detached and curved. The quality of the reduced graphene oxide sheets was evidenced both by X-ray photoelectron spectroscopy and Raman spectroscopy. The electrode material capacitance was determined via electrochemical impedance spectroscopy and cyclic voltammetry. The materials with PEDOT binder had better performance (∼97 F/g) at low operation rates while those with PVDF binder performed better (∼20 F/g) at higher rates, opening up perspectives for their application in supercapacitors.

Structural disorder of graphite and implications for graphite thermometry

Directory of Open Access Journals (Sweden)

M. Kirilova

2018-02-01

Full Text Available Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25  megapascal (MPa and aseismic velocities of 1, 10 and 100 µm s−1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

2012-05-15

As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

Calibration of thin-film dosimeters irradiated with 80-120 kev electrons

DEFF Research Database (Denmark)

Helt-Hansen, J.; Miller, A.; McEwen, M.

2004-01-01

A method for calibration of thin-film dosimeters irradiated with 80-120keV electrons has been developed. The method is based on measurement of dose with a totally absorbing graphite calorimeter, and conversion of dose in the graphite calorimeter to dose in the film dosimeter by Monte Carlo calcul......V electron irradiation. The two calibrations were found to be equal within the estimated uncertainties of +/-10% at 1 s.d. (C) 2004 Elsevier Ltd. All rights reserved....

γ-Irradiation assisted synthesis of graphene oxide sheets supported Ag nanoparticles with single crystalline structure and parabolic distribution from interlamellar limitation

Science.gov (United States)

Yue, Yunhao; Zhou, Baoming; Shi, Jie; Chen, Cheng; Li, Nan; Xu, Zhiwei; Liu, Liangsen; Kuang, Liyun; Ma, Meijun; Fu, Hongjun

2017-05-01

This paper reported a method to fabricate graphene oxide sheets supported Ag nanoparticles (AgNPs/GOS) with single crystalline structure and parabolic distribution without surfactant or functional agent. We used imidazole silver nitrate as intercalation precursor into the layers of graphite oxide, and subsequently reduction and growth of interlamellar AgNPs were induced via γ-irradiation. The results illustrated that the synergism of interlamellar limitation of graphite oxide and fragmentation ability of γ-irradiation could prevent coalescent reaction of AgNPs with other oligomeric clusters, and the single crystalline and small-sized (below 13.9 nm) AgNPs were prepared. Moreover, the content and size of AgNPs exhibited parabolic distribution on GOS surface because the graphite oxide exfoliated to GOS from the edge to the central area of layers. In addition, complete exfoliation degree of GOS and large-sized AgNPs were obtained simultaneously under suitable silver ions concentration. Optimized composites exhibited outstanding surface-enhanced Raman scattering properties for crystal violet with enhancement factor of 1.3 × 106 and detection limit of 1.0 × 10-7 M, indicating that the AgNPs/GOS composites could be applied to trace detection of organic dyes molecules. Therefore, this study presented a strategy for developing GOS supported nanometal with single crystalline structure and parabolic distribution based on γ-irradiation.

Growth kinetics of dislocation loops in irradiated ceramic materials

International Nuclear Information System (INIS)

Ryazanov, A.I.; Kinoshita, C.

2002-01-01

Ceramic materials are expected to be applied in the future fusion reactor as radio frequency (RF) windows, toroidal insulating breaks and diagnostic probes. The radiation resistance of ceramic materials, degradation of the electrical properties and radiation induced conductivity of these materials under neutron irradiation are determined by the kinetics of the accumulation of point defects in the matrix and point defect cluster formation (dislocation loops, voids, etc.). Under irradiation, due to the ionization process, excitation of electronic subsystem and covalent type of interaction between atoms the point defects in ceramic materials are characterized by the charge state (e.g. an F + center, an oxygen vacancy with a single trapped electron) and the effective charge. For the investigation of radiation resistance of ceramic materials for future fusion applications it is very important to understand the physical mechanisms of formation and growth of dislocation loops and voids under irradiation taking into account in this system the effective charge of point defects. In the present paper the physical mechanisms of dislocation loop growth in ceramic material are investigated. For this aim a theoretical model is suggested for the description of the kinetics of point defect accumulation in the matrix taking into account the charge state of the point defects and the effect of an electric field on diffusion migration process of charged point defects. A self-consistent system of kinetic equations describing the generation of electrical fields near dislocation loops and diffusion migration of charged point defects in elastic and electrical fields is formulated. The solution of the kinetic equations allows to find the growth rate of dislocation loops in ceramic materials under irradiation taking into account the charge state of the point defects and the effect of electric and elastic stress fields near dislocation loop on the diffusion processes

AGC-2 Specimen Post Irradiation Data Package Report

Energy Technology Data Exchange (ETDEWEB)

Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-08-01

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

Effect of gamma rays on crystalline materials during irradiation in a reactor

International Nuclear Information System (INIS)

Nikolaenko, V.A.; Karpukhin, V.I.; Gordeev, V.G.

1995-01-01

The article presents and discusses the results of experiments to determine the effect of gamma rays on the change in the properties of diamond, graphite, and structural steel. The materials were irradiated in a VVER type reactor. For diamonds, the effect on the annealing of defects was investigated. As gamma ray intensity increased, the crystal lattice expansion and defect concentration increased. Graphite lattice expansion and the mechanical properties of structural steel were also examined. Graphite lattice expansion increased with increased neutron flux and decreased irradiation temperature. Changes in the impact toughness of structural steel correlated precisely to the gamma ray flux in the experiments. 6 refs., 3 figs

Exfoliated graphite/titanium dioxide nanocomposites for photodegradation of eosin yellow

Energy Technology Data Exchange (ETDEWEB)

Ndlovu, Thabile, E-mail: atkuvarega@gmail.com [University of Swaziland, Department of Chemistry, Private Bag 4, Kwaluseni (Swaziland); Kuvarega, Alex T.; Arotiba, Omotayo A. [University of Johannesburg, Department of Applied Chemistry, P.O. Box 17011, Doornfontein 2028, Johannesburg (South Africa); Sampath, Srinivasan [Indian Institute of Science, Department of Inorganic and Physical Chemistry, Bangalore 560012 (India); Krause, Rui W. [Rhodes University, Department of Chemistry, P.O. Box 94, Grahamstown 6140 South Africa (South Africa); Mamba, Bhekie B., E-mail: bmamba@uj.ac.za [University of Johannesburg, Department of Applied Chemistry, P.O. Box 17011, Doornfontein 2028, Johannesburg (South Africa)

2014-05-01

Graphical abstract: - Highlights: • Preparation of exfoliated graphite (EG) from natural graphite. • Sol–gel anchoring of TiO{sub 2} on exfoliated graphite. • High adsorption and photoactivity was observed for the EG-TiO{sub 2} nanocomposite. • Mechanism of enhancement was proposed. - Abstract: An improved photocatalyst consisting of a nanocomposite of exfoliated graphite and titanium dioxide (EG-TiO{sub 2}) was prepared. SEM and TEM micrographs showed that the spherical TiO{sub 2} nanoparticles were evenly distributed on the surface of the EG sheets. A four times photocatalytic enhancement was observed for this floating nanocomposite compared to TiO{sub 2} and EG alone for the degradation of eosin yellow. For all the materials, the reactions followed first order kinetics where for EG-TiO{sub 2}, the rate constant was much higher than for EG and TiO{sub 2} under visible light irradiation. The enhanced photocatalytic activity of EG-TiO{sub 2} was ascribed to the capability of graphitic layers to accept and transport electrons from the excited TiO{sub 2}, promoting charge separation. This indicates that carbon, a cheap and abundant material, can be a good candidate as an electron attracting reservoir for photocatalytic organic pollutant degradation.

Influence of electron beam irradiation on growth of Phytophthora cinnamomi and its control in substrates

International Nuclear Information System (INIS)

MigdaŁ, Wojciech; Orlikowski, Leszek B.; Ptaszek, Magdalena; Gryczka, Urszula

2012-01-01

Very extensive production procedure, especially in plants growing under covering, require methods, which would allow quick elimination or substantial reduction of populations of specific pathogens without affecting the growth and development of the cultivated plants. Among soil-borne pathogens, the Phytophthora species are especially dangerous for horticultural plants. In this study, irradiation with electron beam was applied to control Phytophthora cinnamomi. The influence of irradiation dose on the reduction of in vitro growth and the population density of the pathogen in treated peat and its mixture with composted pine bark (1:1), as well as the health of Chamaecyparis lawsoniana and Lavandula angustifolia plants were evaluated. Application of irradiation at a dose of 1.5 kGy completely inhibited the in vitro development of P. cinnamomi. This irradiation effect was connected with the disintegration of the hyphae and spores of the species. Irradiation of peat and its mixture with composted pine bark with 10 kGy resulted in the inhibition of stem base rot development in Ch. lawsoniana. Symptoms of the disease were not observed when the substrates were treated with 15 kGy. In the case of L. angustifolia, stem root rot was not observed on cuttings transplanted to infected peat irradiated at a dose of 10 kGy. Irradiation of the horticultural substrates did not affect plant growth. - Highlights: ► Electron beam irradiation is effective against soil-borne pathogens. ► Application of irradiation at dose 1.5 kGy completely inhibited in vitro development of Phytophthora cinnamomi. ► Irradiation of horticultural substrata did not influence the growth of plants.

The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

International Nuclear Information System (INIS)

White, R.J.

2001-01-01

Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

The growth of irradiated candida utilis in molasses media with and without nitrogen

International Nuclear Information System (INIS)

Andini, L.; Harsoyo; Siagian, E.G.

1987-01-01

The growth of irradiated candida utilis in molassea media with and without nitrogen. The experiment has been done to compare the growth of C.utilis on molasses media containing nitrogen and without nitrogen. The C.utilis irradiated with gamma rays from 60 Co (Gamma Cell 220) with different exposure i.e. 0,21, 43, 64, 85 and 106 Gy and then inoculated on 2% molasses medium with and without N, and incubated at 28 0 +- 2 0 C for 0 and 1 day. Parameters observed were the number of colony, the size of colony growth on 2% molasses medium with and without N. The results showed that the growth in media molasses with N was different from that without N. The incubation effect seemed to be a significant, while the effect of irradiation was not significant, which was probably due to the insufficient dose used for inducing stimulation. The interaction between the added Nand incubation was also significant. 13 refs

Cranial irradiation of young rats impairs later learning and growth

International Nuclear Information System (INIS)

Overmier, J.B.; Carroll, M.E.; Patten, R.; Krivit, W.; Kim, T.H.

1979-01-01

Young rats (26 days) were exposed to ionizing radiation of the head of 0, 1200, 2400 or 3000 rads total in 200 rads/day doses. The subsequent growth of irradiated rats was permanently impaired: such impairment was positively related to amount of irradiation. Beginning in adolescence, rats were trained on a horizontal/vertical visual discrimination in a runway task and although all four groups mastered the discrimination, they differed in their patterns of acquisition. These results indicated long term effects and are associated with a cranial irradiation regimen similar to that given to children suffering acute lymphocytic leukemia (ALL). (author)

Inactivation of the Escherichia coli chromosome during growth after ultraviolet irradiation

Energy Technology Data Exchange (ETDEWEB)

Medic-Petranovic, M; Trgovcevic, Z; Novak, D; Petranovic, D [Institut Rudjer Boskovic, Zagreb (Yugoslavia)

1977-07-01

Cells of the repair-proficient E.coli AB1157 strain and its lysogenic E.coli AB1157 (lambda c1857 ind) counterpart have been uv irradiated in an attempt to define when and why cells that are destined to die reach their biological end-point in the course of post-irradiation incubation. The thermo-inducibility of the lambdac1857 ind lysogens was first determined, since this reflects the functional integrity of the pro-viral part of the bacterial chromosome and that of the bacterial cytoplasm. The capacity (i.e. the ability of the irradiated cells to support growth of the unirradiated phage) was then determined, since this depends on the functional integrity of the bacterial cytoplasm. A progressive decrease in the ability of the lysogens to be heat-induced always preceded the decrease in capacity for the phage growth. The results strongly suggest that wild-type E.coli cells destined to die after exposure to moderate doses of uv-light reach their end point within 4 hours of post-irradiation incubation, probably as a result of the functional failure of the whole chromosome.

Enhanced oxidation resistance of SiC coating on Graphite by crack healing at the elevated temperature

Energy Technology Data Exchange (ETDEWEB)

Park, Jae-Won; Kim, Eung-Seon; Kim, Jae-Un; Kim, Yoo-Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Windes, William E. [Idaho National Laboratory, Idaho (United States)

2015-10-15

An oxidation protective SiC coating on the graphite components could assist in slowing the oxidation down. However, the irradiation induced dimensional changes in the graphite (shrinkage followed by swelling) can occur, while the SiC CVD coating has been reported to swell even at a low dose neutron irradiation. In this work, functionally gradient electron beam evaporative coating with an ion beam processing was firstly conducted and then SiC coating on the FG coating to the desired thickness is followed. For the crack healing, both the repeated EB-PVD and CVD were performed. Oxidation and thermal cycling tests of the coated specimens were performed and reflected in the process development. In this work, efforts have been paid to heal the cracks in the SiC coated layer on graphite with both EB-PVD and CVD. CVD seems to be more appropriate coating method for crack healing probably due to its excellent crack-line filling capability for high density and high aspect ratio.

The enhanced effects of antibiotics irradiated of extremely high frequency electromagnetic field on Escherichia coli growth properties.

Science.gov (United States)

Torgomyan, Heghine; Trchounian, Armen

2015-01-01

The effects of extremely high frequency electromagnetic irradiation and antibiotics on Escherichia coli can create new opportunities for applications in different areas—medicine, agriculture, and food industry. Previously was shown that irradiated bacterial sensitivity against antibiotics was changed. In this work, it was presented the results that irradiation of antibiotics and then adding into growth medium was more effective compared with non-irradiated antibiotics bactericidal action. The selected antibiotics (tetracycline, kanamycin, chloramphenicol, and ceftriaxone) were from different groups. Antibiotics irradiation was performed with low intensity 53 GHz frequency during 1 h. The E. coli growth properties—lag-phase duration and specific growth rate—were markedly changed. Enhanced bacterial sensitivity to irradiated antibiotics is similar to the effects of antibiotics of higher concentrations.

Irradiation-induced growth of zircaloy and its effects on the mechanical design of fuel assemblies

International Nuclear Information System (INIS)

Yao Pu

1991-01-01

Zircaloy growth could be induced due to irradiation. The ammount of growth is described as a function of texture, irradiation temperature, fast neutron fluence and the reduction of cold work, and it should be given great attention in the mechanical design of fuel assemblies

Laser surface graphitization to control friction of diamond-like carbon coatings

Science.gov (United States)

Komlenok, Maxim S.; Kononenko, Vitaly V.; Zavedeev, Evgeny V.; Frolov, Vadim D.; Arutyunyan, Natalia R.; Chouprik, Anastasia A.; Baturin, Andrey S.; Scheibe, Hans-Joachim; Shupegin, Mikhail L.; Pimenov, Sergei M.

2015-11-01

To study the role of laser surface graphitization in the friction behavior of laser-patterned diamond-like carbon (DLC) films, we apply the scanning probe microscopy (SPM) in the lateral force mode (LFM) which allows to obtain simultaneously the lateral force and topography images and to determine local friction levels in laser-irradiated and original surface areas. Based on this approach in the paper, we report on (1) laser surface microstructuring of hydrogenated a-C:H and hydrogen-free ta-C films in the regime of surface graphitization using UV laser pulses of 20-ns duration and (2) correlation between the structure and friction properties of the laser-patterned DLC surface on micro/nanoscale using SPM/LFM technique. The SPM/LFM data obtained for the surface relief gratings of graphitized microstructures have evidenced lower friction forces in the laser-graphitized regions. For the hydrogenated DLC films, the reversible frictional behavior of the laser-graphitized micropatterns is found to take place during LFM imaging at different temperatures (20 and 120 °C) in ambient air. It is revealed that the lateral force distribution in the laser-graphitized areas is shifted to higher friction levels (relative to that of the unirradiated surface) at temperature 120 °C and returned back to the lower friction during the sample cooling to 20 °C, thus confirming an influence of adsorbed water layers on the nanofriction properties of laser-graphitized micropatterns on the film surface.

Draft of standard for graphite core components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

2010-01-01

For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)