Encapsulating of high-level radioactive waste with use of pyrocarbon and silicon carbide coatings

International Nuclear Information System (INIS)

Chernikov, A.

2007-01-01

It is known that high-level radioactive waste (HLW) constitute a real danger to biosphere, especially that their part, which contains transuranium and long-lived radionuclides resulting during reprocessing of nuclear fuel industrial and power reactors. Such waste contains approximately 99 % of long-lived fission products and transplutonium elements. At present, the concept of multi barrier protection of biosphere from radioactive waste is generally acknowledged. The main barriers are the physicochemical form of waste and enclosing strata of geological formation at places of waste-disposal. Applied methods of solidification of HLW with preparation of phosphatic and borosilicate glasses do not guarantee in full measure safety of places of waste-disposal of solidified waste in geological formations during thousand years. One promising way to improve HLW handling safety is placing of radionuclides in mineral-like matrixes similar to natural materials. The other possible way to increase safety of HLW disposal places is suggested for research by experts of Russian research institutes, for example, in the proposal for the Project of ISTC and considered in the present report, is to introduce an additional barrier on a radionuclides migration path by coating of HLW particles. Unique protective properties of pyrocarbon and silicon carbide such as low coefficients of diffusion of gaseous and solid fission products and high chemical and radiation stability [1] attract attention to these materials for coating of solidified HLW. The objective of the Project is the development of method of HLW encapsulating with use of pyrocarbon and silicon carbide coatings. To gain this end main direction of researches, including analysis of various encapsulation processes of fractionated HLW, and expected results are presented. Realization of the Project will allow to prove experimentally the efficiency of the proposed approach in the solution of the problem of HLW conditioning and ecological

Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

Energy Technology Data Exchange (ETDEWEB)

Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

2008-07-01

Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

Interaction of Al2O3xSiO2 alloyed uranium oxide with pyrocarbon coating of fuel particles under irradiation

International Nuclear Information System (INIS)

Chernikov, A.S.; Khromov, Yu.F.; Svistunov, D.E.; Chujko, E.E.

1989-01-01

Method of comparative data analysis for P O2 and P CO was used to consider interaction in fuel particle between pyrocarbon coating and fuel sample, alloyed with alumosilicate addition. Equations of interaction reactions for the case of hermetic and depressurized fuel particle are presented. Calculations of required xAl 2 O 3 XySiO 2 content, depending on oxide fuel burnup, were conducted. It was suggested to use silicon carbide for limitation of the upper level of CO pressure in fuel particle. Estimation of thermal stability of alumosilicates under conditions of uranium oxide burnup equals 1100 and 1500 deg C for Al/Si ratio in addition 1/1 and 4/1 respectively

X-Ray Researches GF Siliconized Materials on Pyrocarbon Sheaf and on the Basis of Graphite of Mark EG-0

International Nuclear Information System (INIS)

Gurin, V.A.; Gurin, I.V.; Kovtun, G.P.; Malykhin, D.G.; Bukolov, A.N.

2005-01-01

A methodological addition to a quantitative analysis of binary phase structure of materials on measurements of X-ray lines intensities worked out conformably to research of siliconized graphitic materials. Distinctions in X-rays absorption factors of phase components at a various degree of phases mixture are taken into account. An apparatus of the probability theory is applied. A parameter of mixture degree of phases is submitted as a specific area size of interphase. Quantitative X-ray researches of a phase structure of siliconized materials are carried out on the basis of carbon fabrics and graphitic powders; both were sheafed by pyrocarbon. In examined samples structures C-SiC and SiC-Si were obtained. The correlation of the phase structure of materials with the apparent density of the initial carbon basis is seen. The opportunity of a practical obtaining of materials with the host degree of their siliconizing is confirmed

Detection and control of as-produced pyrocarbon permeability in biso-coated high-temperature gas-cooled reactor fuel particles

International Nuclear Information System (INIS)

Stinton, D.P.; Thiele, B.A.; Lackey, W.J.; Morgan, C.S.

1980-05-01

About 60 Biso-coated particle batches with coatings deposited in either 0.13- or 0.24-m dia coaters were studied in this work. These batches were carefully characterized for permeability by neon-helium intrusion, long-term chlorination followed by radiography, and fission gas release. These methods of permeability measurement were compared and correlated with deposition conditions as well as pyrocarbon properties. The results from several irradiation tests were also used to evaluate the validity of the permeability measurements. The neon-helium and long-term chlorination techniques correlated very clearly. Coatings with neon-to-helium ratios below 0.3 were gastight by the chlorination procedure, whereas those with ratios above 0.4 were permeable. The fission gas release technique was unable to distinguish between slightly permeable coatings and gastight ones. Therefore, neon-helium and long-term chlorination procedures are preferred over the fission gas release technique. Results from several irradiation tests verified that coatings with neon-to-helium ratios below 0.3 were gastight, whereas those with ratios above about 0.4 were permeable. 10 figures, 2 tables

Influence of the matrix texture on the fracture behavior of 2D carbon/carbon composites

International Nuclear Information System (INIS)

Xu Guozhong; Li Hejun; Bai Ruicheng; Wei Jian; Zha, Yanqiang

2008-01-01

The influence of matrix texture on the fracture behavior of 2D carbon/carbon composites infiltrated by isobaric, isothermal CVI technique at ambient pressure was investigated. The flexural strength of the as-obtained samples has been studied using a three-point bending test. After flexural test, the texture of pyrocarbon on the fracture surface and the morphology of the fracture surface were observed by polarized light microscopy and scanning electron microscopy, respectively. The results show that the sample with pure medium-textured pyrocarbon exhibits typical brittle fracture behavior due to no sliding between sub-layers in the medium-textured pyrocarbon layer. However, both the sample with the three-layer structure of medium-textured pyrocarbon, high-textured pyrocarbon and low-textured pyrocarbon, and the sample with the double-layer structure of medium-textured pyrocarbon and high-textured pyrocarbon exhibit remarkable pseudo-plastic fracture behavior, which is caused by the sliding occurred between different textured pyrocarbon layers and between sub-layers in high-textured pyrocarbon layer

Structures and electrochemical properties of pyrolytic carbon films infiltrated from gas phase into electro-conductive substrates derived from wood

International Nuclear Information System (INIS)

Ohzawa, Yoshimi; Mitani, Masami; Li, Jianling; Nakajima, Tsuyoshi

2004-01-01

Using the pressure-pulsed chemical vapor infiltration technique, pyrolytic carbon (pyrocarbon) films were deposited into two sorts of conductive porous substrates, that is, the carbonized wood (A) and the TiN-coated wood (B). Structures and electrochemical properties were investigated as the negative electrodes of lithium-ion secondary battery. The electrodes had the three-dimensionally continuous current paths in the pyrocarbon-based anodes without the organic binders and the additional conductive fillers. The pyrocarbon films adhered tightly to the carbonized wood or TiN as current collector. These macro-structures of electrodes were effective in improving the high rate property. The sort of substrates affected the nano-structure of pyrocarbon. The pyrocarbon in sample (A) had the relatively high crystallinity, whereas the pyrocarbon in sample (B) was disordered. The capacity of pyrocarbon in sample (B) was higher than that of sample (A), reflecting the disordered microstructure of pyrocarbon film (B). However, sample (A) showed higher Coulombic efficiency at first cycle (i.e. 87%) than that of sample (B), which would result from the high crystallinity, laminar microstructure and low surface area of pyrocarbon in sample (A)

Methods for the characterization of pyrolytic deposited carbon

International Nuclear Information System (INIS)

Bongartz, K.; Hoven, H.; Koizlik, K.

Pyrocarbon is deposited as a coating material on fuel kernels used in HTGRs. For the development of particle coatings specified for various reactor designs, it is necessary to know the properties of pyrocarbon and their changes by neutron irradiation. In this report, procedures are described which are used to characterize pyrocarbon: measurement of geometry, density, microporosity, apparent crystallite size, anisotropy of orientation, modulus of elasticity, and strength of coatings, as well as ceramography, etching by oxidation, secondary and transmission electron microscopy. (auth)

Spectral ellipsometry of nanodiamond composite

International Nuclear Information System (INIS)

Yastrebov, S.G.; Ivanov-Omskij, V.I.; Gordeev, S.K.; Garriga, M.; Alonso, I.A.

2006-01-01

Methods of spectral ellipsometry were applied for analysis of optical properties of nanodiamond based composite in spectral region 1.4-5 eV. The nanocomposite was synthesized by molding of ultradispersed nanodiamond powder in the course of heterogeneous chemical reaction of decomposition of methane, forming pyrocarbon interconnecting nanodiamond grains. The energy of σ + π plasmon of pyrocarbon component of nanodiamond composite was restored which proves to be ∼ 24 eV; using this value, an estimation was done of pyrocarbon matrix density, which occurs to be 2 g/cm 3 [ru

CVD in nuclear energy

International Nuclear Information System (INIS)

Nickel, H.

1981-08-01

CVD-deposited pyrocarbon, especially the coatings of nuclear fuel kernels show a structure depending on many parameters such as deposition temperature, nature and pressure of the pyrolysis gas, nature of the substrate, geometry of the deposition system, etc. Because of the variety of pyrocarbon different characterization methods have been developed or qualified for this new application. Additionally classical characterization procedures are available. Beside theoretical aspects concerning the formation and deposition mechanism of pyrocarbon from the gas phase the behaviour of such coatings under irradiation with fast neutrons is discussed. (orig.) [de

Hoogsteen base pairs proximal and distal to echinomycin binding sites on DNA

International Nuclear Information System (INIS)

Mendel, D.; Dervan, P.B.

1987-01-01

Forms of the DNA double helix containing non-Watson-Crick base-pairing have been discovered recently based on x-ray diffraction analysis of quionoxaline antibiotic-oligonucleotide complexes. In an effort to find evidence for Hoogsteen base-pairing at quinoxaline-binding sites in solution, chemical footprinting (differential cleavage reactivity) of echinomycin bound to DNA restriction fragments was examined. The authors report that purines (A>G) in the first and/or fourth base-pair positions of occupied echinomycin-binding sites are hyperreactive to diethyl pyrocarbonate. The correspondence of the solid-state data and the sites of diethyl pyrocarbonate hyperreactivity suggests that diethyl pyrocarbonate may be a sensitive reagent for the detection of Hoogsteen base-pairing in solution. Moreover, a 12-base-pair segment of alternating A-T DNA, which is 6 base pairs away from the nearest strong echinomycin-binding site, is also hyperreactive to diethyl pyrocarbonate in the presence of echinomycin. This hyperreactive segment may be an altered form of right-handed DNA that is entirely Hoogsteen base-paired

Progress in Studies on Carbon and Silicon Carbide Nanocomposite Materials

International Nuclear Information System (INIS)

Xiao, P.; Chen, J.; Xian-feng, X.

2010-01-01

Silicon carbide nanofiber and carbon nanotubes are introduced. The structure and application of nanotubers (nanofibers) in carbon/carbon composites are emphatically presented. Due to the unique structure of nanotubers (nanofibers), they can modify the microstructure of pyrocarbon and induce the deposition of pyrocarbon with high text in carbon/carbon composites. So the carbon/carbon composites modified by CNT/CNF have more excellent properties.

The behaviour of spherical HTR fuel elements under accident conditions

Energy Technology Data Exchange (ETDEWEB)

Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

1985-07-01

Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

Operation and postirradiation examination of ORR capsule OF-2: accelerated testing of HTGR fuel

International Nuclear Information System (INIS)

Tiegs, T.N.; Thoms, K.R.

1979-03-01

Irradiation capsule OF-2 was a test of High-Temperature Gas-Cooled Reactor fuel types under accelerated irradiation conditions in the Oak Ridge Research Reactor. The results showed good irradiation performance of Triso-coated weak-acid-resin fissile particles and Biso-coated fertile particles. These particles had been coated by a fritted gas distributor in the 0.13-m-diam furnace. Fast-neutron damage (E > 0.18 MeV) and matrix-particle interaction caused the outer pyrocarbon coating on the Triso-coated particles to fail. Such failure depended on the optical anisotropy, density, and open porosity of the outer pyrocarbon coating, as well as on the coke yield of the matrix. Irradiation of specimens with values outside prescribed limits for these properties increased the failure rate of their outer pyrocarbon coating. Good irradiation performance was observed for weak-acid-resin particles with conversions in the range from 15 to 75% UC 2

Boron-bearing species in ceramic matrix composites for long-term aerospace applications

International Nuclear Information System (INIS)

Naslain, R.; Guette, A.; Rebillat, F.; Pailler, R.; Langlais, F.; Bourrat, X.

2004-01-01

Boron-bearing refractory species are introduced in non-oxide ceramic matrix fibrous composites (such as SiC/SiC composites) to improve their oxidation resistance under load at high temperatures with a view to applications in the aerospace field. B-doped pyrocarbon and hex-BN have been successfully used as interphase (instead of pure pyrocarbon) either as homogeneous or multilayered fiber coatings, to arrest and deflect matrix cracks formed under load (mechanical fuse function) and to give toughness to the materials. A self-healing multilayered matrix is designed and used in a model composite, which combines B-doped pyrocarbon mechanical fuse layers and B- and Si-bearing compound (namely B 4 C and SiC) layers forming B 2 O 3 -based fluid healing phases when exposed to an oxidizing atmosphere. All the materials are deposited by chemical vapor infiltration. Lifetimes under tensile loading of several hundreds hours at high temperatures are reported

Spectral ellipsometry of a nanodiamond composite

International Nuclear Information System (INIS)

Yastrebov, S. G.; Gordeev, S. K.; Garriga, M.; Alonso, I. A.; Ivanov-Omskii, V. I.

2006-01-01

Optical properties of a nanodiamond composite were analyzed by methods of spectral ellipsometry in the range of photon energies 1.4-5 eV, which are characteristic of π-π* transitions in amorphous carbon. The nanocomposite was synthesized by molding nanodiamond powder with subsequent binding of diamond nanoparticles by pyrocarbon formed as a result of the heterogeneous chemical reaction of methane decomposition. The dispersion curves of the imaginary and real parts of the dielectric function were reconstructed. It is shown that the imaginary part of the dielectric function can be represented as the sum of two components generated by the two types of π-π* optical transitions. The maximum contribution of the transitions of the first and second types manifests itself at energies of 2.6 and 5.6 eV, respectively, which correspond to peaks in optical density at 2.9 and 6.11 eV. It was established that the main specific features of the normalized optical density of the nanodiamond composite almost coincide with those for poly(para-phenylenevinylene). It was found that the energy of a σ + π plasmon of the pyrocarbon component of the nanodiamond composite is 24.2 eV. On the basis on this value, the pyrocarbon density matrix was estimated to be 2 g/cm 3 . Within the concepts of optimum filling of an elementary volume by carbon atoms in an amorphous material with such a density, the allotropic composition of the pyrocarbon matrix was restored

Irradiation-induced dimensional changes of poorly crystalline carbons

International Nuclear Information System (INIS)

Bullock, R.E.

1979-01-01

Data are presented on irradiation-induced changes of poorly crystalline carbons at high temperatures(>900 0 C). The materials surveyed include: (1) carbon fibers, (2) glassy carbons, (3) carbonaceous matrix materials for HTGR fuel rods and (4) pyrocarbons. The materials are listed in order of increasing stability, with maximum strains ranging from more than 50% for fibers to less than 10% for pyrocarbons. Dimensional changes of highly anisotropic carbon fibers appear to be sensitive to irradiation temperature, as slightly anisotropic pyrocarbons are, whereas temperature seems to have little influence on the behavior of isotropic glassy carbons over the range from 600 to 1350 0 C. Dimensional changes for graphite-filled matrix materials were roughly isotropic on the average and did not seem to be strongly temperature dependent for the lower fluences investigated. Increased graphite filler lowered volumetric dimensional changes of the matrix in agreement with a rule-of-mixtures relationship between change components for the filler and the less-stable binder phases. Instabilities of all of the poorly crystalline materials were generally greater than those for more crystalline carbons under the same conditions, including highly orientated graphites that approximate single-crystal behavior. (author)

On the fractography of overload, stress corrosion, and cyclic fatigue failures in pyrolytic-carbon materials used in prosthetic heart-valve devices.

Science.gov (United States)

Ritchie, R O; Dauskardt, R H; Pennisi, F J

1992-01-01

A scanning electron microscopy study is reported of the nature and morphology of fracture surfaces in pyrocarbons commonly used for the manufacture of mechanical heart-valve prostheses. Specifically, silicon-alloyed low-temperature-isotropic (LTI)-pyrolytic carbon is examined, both as a coating on graphite and as a monolithic material, following overload, stress corrosion (static fatigue), and cyclic fatigue failures in a simulated physiological environment of 37 degrees C Ringer's solution. It is found that, in contrast to most metallic materials yet in keeping with many ceramics, there are no distinct fracture morphologies in pyro-carbons which are characteristic of a specific mode of loading; fracture surfaces appear to be identical for both catastrophic and subcritical crack growth under either sustained or cyclic loading. We conclude that caution should be used in assigning the likely cause of failure of pyrolytic carbon heart-valve components using fractographic examination.

Compilation of reports presented by the IRW together with participating industrial firms at the Reactor Session 1977 of the German Atomic Forum e.V./KTG (29 March--1 April 1977, Mannheim)

Energy Technology Data Exchange (ETDEWEB)

Rottmann, J. [comp.

1977-02-15

Separate abstracts were prepared for the 8 reports on problematics of HTR reactors: ¹¹⁰Ag retention; pyrocarbon coatings on fuel particles; SiC corrosion; irradiation effects on claddings, spherical fuel elements, and graphite; quality control; and high-temperature alloy testing in HTR He. (DLC)

Coating of waste containing ceramic granules

International Nuclear Information System (INIS)

Neumann, W.; Kofler, O.

1979-01-01

Simulated high-level waste granules produced by fluidized-bed calcination were overcoated by chemical vapor deposition (CVD) with pyrocarbon and nickel in laboratory-scale experiments. Successful development enables pyrocrbon deposition at temperatures of 600 to 800 0 K. The coated granules have excellent properties for long-term waste storage

Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

International Nuclear Information System (INIS)

Li Junfeng

2016-01-01

High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix

1976 scientific progress report. [Fuel and coating materials for HTGR]; Wissenschaftlicher Ergebnisberict 1976

Energy Technology Data Exchange (ETDEWEB)

Nickel, H.

1976-07-01

Activities at the Institute for Reactor Materials in the production and properties of high temperature gas cooled reactor fuel and coating materials are summarized. Major emphasis was placed on investigations of pyrocarbon, BISO and TRISO coatings, uranium and thorium oxides and carbides, and graphite and matrix materials. A list of publications is included. (HDR)

Transfer of fissile material through shielding coatings in emergency heating of HTGR coated particles

International Nuclear Information System (INIS)

Gudkov, A.N.; Zhuravkov, S.G.; Koptev, M.A.; Kurepin, A.D.

1990-01-01

The measurement results of leakage dynamics of fissile material from the coated particles within a temperature range of 1200 + 2000 deg. C are given. The methods of carrying out the experiments are briefly described. The relation of the leakage rate of uranium-235 from CP (coated particles) with the pyrocarbonic coatings has been obtained. (author)

Irradiation performance of HTGR fuel in HFIR capsule HT-31

International Nuclear Information System (INIS)

Tiegs, T.N.; Robbins, J.M.; Hamner, R.L.; Montgomery, B.H.; Kania, M.J.; Lindemer, T.B.; Morgan, C.S.

1979-05-01

The capsule was irradiated in the High Flux Isotope Reactor at ORNL to peak particle temperatures up to 1600 0 C, fast neutron fluences (0.18 MeV) up to 9 x 10 25 n/m 2 , and burnups up to 8.9% FIMA for ThO 2 particles. The oxygen release from plutonium fissions was less than calculated, possibly because of the solid solution of SrO and rare earth oxides in UO 2 . Tentative results show that pyrocarbon permeability decreases with increasing fast neutron fluence. Fission products in sol-gel UO 2 particles containing natural uranium mostly behaved similarly to those in particles containing highly enriched uranium (HEU). Thus, much of the data base collected on HEU fuel can be applied to low-enriched fuel. Fission product palladium penetrated into the SiC on Triso-coated particles. Also the SiC coating provided some retention of /sup 110m/Ag. Irradiation above about 1200 0 C without an outer pyrocarbon coating degraded the SiC coating on Triso-coated particles

Contribution to the study of hard, low-density pyrolytic carbons; Contribution a l'etude des carbones pyrolytiques de variete dure et de faible densite

Energy Technology Data Exchange (ETDEWEB)

Boutin, F R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

Apparent contradictions in the properties of pyrolytic carbons obtained at 1600 deg C (hardness and graphitization) are studied. It is shown that structure of the deposit is turbostratic with high internal stresses ({delta}{sup -2}), and it graphitizes (by thermal treatment over 2000 deg C) in a similar manner to graphitisable carbon. Because the deposit forms lamellar compounds, it is presumed that the structure is similar to that of graphitisable carbon. Since it is not structure dependant, the hardness originates from the 'growth texture' and is not comparable with the hardness of a non-graphitisable carbon. The pyrolytic carbon studied is composed of regions, on the overage a few microns across, formed by the stacking of small carbon platelets, interlocked and showing a preferred orientation. The mis-orientation of the various regions produces general disorientation. We estimate that the introduction of the particles of some material such as thermal black which are observed in the electron microscope are responsible for the mis-orientation. The density and hardness of the deposit are a result of the interlocking of platelets, which creates a closed porosity and prevents any sliding of the atomic planes. (author) [French] On etudie les proprietes apparemment contradictoires du pyrocarbone 1600 deg C, durete et graphitabilite. On montre que le pyrocarbone possede la structure d'un carbone turbostratique a fort taux de distorsion et qu'il subit, par traitement a des temperatures superieures a 2000 deg C, une transformation de graphitation comparable a celle que l'on observe sur les cokes graphitables. Comme le pyrocarbone forme de plus des composes d'insertion, sa structure est comparable a celle d'un carbone graphitable. La durete, qui n'a pas d'origine structurale, est donc liee a la 'texture de croissance' du depot et ne peut etre comparee a celle d'un coke dur et non graphitable. Le pyrocarbone etudie est constitue de domaines, dont les dimensions sont de l

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)

High level waste containing granules coated and embedded in metal as an alternative to HLW glasses

International Nuclear Information System (INIS)

Neumann, W.

1980-01-01

Simulated high level waste containing granules were overcoated with pyrocarbon or nickel respectively. The coatings were performed by the use of chemical vapour deposition in a fluidized bed. The coated granules were embedded in an aluminium-silicon-alloy to improve the dissipation of radiation induced heat. The metal-granules-composites obtained were of improved product stability related to the high level waste containing glasses. (orig.) [de

Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

International Nuclear Information System (INIS)

Nickel, H.

1990-05-01

The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

Ceramic matrix micro-composites prepared by P-Rcvd within the (Ti-Si-B-C) system

International Nuclear Information System (INIS)

Jacques, Sylvain

2014-01-01

Nano-scale carbide multilayered inter-phases were deposited within the (Ti-Si-B-C) system by pressure-Pulsed Reactive Chemical Vapor Deposition (P-RCVD) on single filament Hi-Nicalon fibers and embedded in a SiC matrix sheath. The Reactive method, in which the titanium-containing layer growth involves the consumption of the pre-deposited Si-B-C sublayer, allowed TiC- and TiB 2 -based films to be obtained with a porous multilayer microstructure as a result of the Kirkendall effect. A first difficulty relied on the protection of the fiber surface which was very sensitive to chemical attack by P-RCVD. This difficulty could be circumvented through a first deposited SiC sub-layer thick enough to protect the surface of the fiber. But, because the porosity volume fraction was still not high enough, the role of mechanical fuse of these pyrocarbon-free inter-phases could not be evidenced from the micro-composite tensile curves, which remained fully linear up to the failure. Finally, the P-RCVD method was applied to the matrix processing itself. Micro-composites, this time with a pyrocarbon interphase but also with new matrix materials such as Ti 3 SiC 2 , were prepared and characterized. (author)

Performance of HTGR fuel in HFIR capsule HT-33

International Nuclear Information System (INIS)

Tiegs, T.N.; Robbins, J.M.

1979-06-01

Irradiation capsule HT-33 was a cooperative effort between General Atomic Company (GA) and Oak Ridge National Laboratory (ORNL). In this capsule ThO 2 particles (fabricated by GA), low-enriched uranium particles, inert carbon particles, and various fuel rod matrices were tested under accelerated irradiation in the High-Flux Isotope Reactor. Visual examination showed good irradiation behavior for fuel rods with slug-injected matrices (using a pitch binder) and warm-molded matrices (using a thermosetting resin binder). Rod debonding improved somewhat with fuel rods that used GLCC H-451 ground graphite shim particles rather than Speer fluid coke shim particles. Measurements of permeability (by inert gas intrusion) of the pyrocarbon on the inert particles showed that the disorder created by the neutron flux did not increase the inert gas permeability. Metallographic examination of Triso-coated particles irradiated both with and without an outer pyrocarbon coating revealed that the outer coating is necessary to suppress SiC degradation at temperatures above approximately 1375 0 C. The fission product behavior (determined by the electron microprobe) was similar in both low-enriched and high-enriched uranium particles made from weak-acid resins. Furthermore, fission product palladium caused severe SiC corrosion at time-averaged temperatures above 1400 0 C

Annual Report 2013-2014: Theoretical Studies of Nerve Agents Adsorbed on Surfaces

Science.gov (United States)

2014-07-08

magnesium oxide. Journal of Physical Chemistry B 2004, 108, 5294-5303. 7. Michalkova, A.; Paukku, Y.; Majumdar, D.; Leszczynski, J., Theoretical study...M.; Hyre, A. M., Ultraviolet Raman Spectra and Cross-Sections of the G-series Nerve Agents. Applied Spectroscopy 2008, 62, 1078-1083. 12. DaBell...mol) 24 2.3 Raman and DFT study of pyrocarbonate 2.3.1 In-situ Raman spectroscopic investigation The Li2CO3 and Na2CO3 eutectic mixture (52:48

Graphical analysis of processes with multiple activation energies

International Nuclear Information System (INIS)

Lachter, J.; Bragg, R.H.; Close, E.

1986-01-01

The activation energies characterizing a kinetic process are derived from the slopes of the Arrhenius diagrams obtained by plotting rate constants versus reciprocal temperature. Those rate constants correspond to the shifts along the time axis needed to superpose the successive isotherms. A general method based on Chebyshev interpolation is proposed for the optimization of the superposition of the experimental data points. This method is applied to determine the activation energies of the graphitization kinetics of the interlayer spacings of pitch coke and pyrocarbon samples

In-pile tests of HTGR fuel particles and fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

1985-01-01

Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

Development of a pneumatic transfer system for HTGR recycle fuel particles

International Nuclear Information System (INIS)

Mack, J.E.; Johnson, D.R.

1978-02-01

In support of the High-Temperature Gas-Cooled Reactor (HTGR) Fuel Refabrication Development Program, an experimental pneumatic transfer system was constructed to determine the feasibility of pneumatically conveying pyrocarbon-coated fuel particles of Triso and Biso designs. Tests were conducted with these particles in each of their nonpyrophoric forms to determine pressure drops, particle velocities, and gas flow requirements during pneumatic transfer as well as to evaluate particle wear and breakage. Results indicated that the material can be pneumatically conveyed at low pressures without excessive damage to the particles or their coatings

Status of the fuel stress and failure rate calculations at KFA

International Nuclear Information System (INIS)

Bongartz, K.

1980-11-01

In this report a new model for calculating stresses in the SiC layer of TRISO coated particles is presented. The gain in computer time with respect to the Walther model used up to now is a factor of 100. The restrictions of this model are - application is only possible to TRISO particles and not BISO which can be handled with the Walther model as well, the SiC layer is regarded as rigid: in fact, its Young's modulus is higher by a factor of 10 as compared to that of the Pyrocarbon layers. (orig.) [de

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

Silver release from coated particle fuel

International Nuclear Information System (INIS)

Brown, P.E.; Nabielek, H.

1977-03-01

The fission product Ag-110 m released from coated particles can be the dominant source of radioactivity from the core of a high temperature reactor in the early stages of the reactor life and possibly limits the accessability of primary circuit components. It can be shown that silver is retained in oxide fuel by a diffusion process (but not in carbide or carbon-diluted fuel) and that silver is released through all types of pyrocarbon layers. The retention in TRISO particles is variable and seems to be mainly connected with operating temperature and silicon carbide quality. (orig.) [de

Review on characterization methods applied to HTR-fuel element components

International Nuclear Information System (INIS)

Koizlik, K.

1976-02-01

One of the difficulties which on the whole are of no special scientific interest, but which bear a lot of technical problems for the development and production of HTR fuel elements is the proper characterization of the element and its components. Consequently a lot of work has been done during the past years to develop characterization procedures for the fuel, the fuel kernel, the pyrocarbon for the coatings, the matrix and graphite and their components binder and filler. This paper tries to give a status report on characterization procedures which are applied to HTR fuel in KFA and cooperating institutions. (orig.) [de

Cesium transport data for HTGR systems

International Nuclear Information System (INIS)

Myers, B.F.; Bell, W.E.

1979-09-01

Cesium transport data on the release of cesium from HTGR fuel elements are reviewed and discussed. The data available through 1976 are treated. Equations, parameters, and associated variances describing the data are presented. The equations and parameters are in forms suitable for use in computer codes used to calculate the release of metallic fission products from HTGR fuel elements into the primary circuit. The data cover the following processes: (1) diffusion of cesium in fuel kernels and pyrocarbon, (2) sorption of cesium on fuel rod matrix material and on graphite, and (3) migration of cesium in graphite. The data are being confirmed and extended through work in progress

Dielectric Properties of SiCf/PyC/SiC Composites After Oxidation

Institute of Scientific and Technical Information of China (English)

SONG Huihui; ZHOU Wancheng; LUO Fa; QING Yuchang; CHEN Malin; LI Zhimin

2016-01-01

In this paper, the SiC fiber-reinforced SiC matrix composites with a 0.15mm thick pyrocarbon interphase (notedas SiCf/PyC/SiC) were prepared by chemical vapor infiltration (CVI). The SiCf/PyC/SiC were oxidized in air at 950℃ for 50h. The dielectric properties after this high temperature oxidation were investigated in X-band from room temperature (RT) to 700℃. Results suggested that:e' of the SiCf/PyC/SiC after oxidation increased at first then de-creased with temperature elevating;e" increased with temperature raising in the temperature range studied.

The properties of spherical fuel elements and its behavior in the modular HTR

International Nuclear Information System (INIS)

Lohnert, G.H.; Ragoss, H.

1985-01-01

The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO 2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

Physicochemical analysis of interaction of oxide fuel with pyrocarbon coatings of fuel particles

International Nuclear Information System (INIS)

Lyutikov, R.A.; Khromov, Yu.F.; Chernikov, A.S.

1990-01-01

Equilibrium pressure of (CO+Kr,Xe) gases inside fuel particle with oxide kern depending on design features of fuel particle, on temperature. on (O/U) initial composition and fuel burnup is calculated using the suggested model. Analysis of possibility for gas pressure reduction by means of uranium carbide alloying of kern and degree increase of solid fission product retention (Cs for example) during alumosilicate alloying of uranium oxide is conducted

On the kinetics of high-temperature interaction of tungsten with light hydrocarbons

International Nuclear Information System (INIS)

Kharatyan, S.L.; Chatilyan, A.A.; Merzhanov, A.G.

1989-01-01

Comparative investigation of tungsten carbidizing treatment in ethylene, acetylene and methane media at T=1750-2500 deg C and p=2-10 Torr is carried out by the electrothermographical method. In all cases interaction is shown to proceed in stages due to step-by-step formation of carbide phases of tungsten W 2 C and WC as well as pyrocarbon. It is established that efficiency of carbidizing treatment is turned out to be maximum in methane medium in spite of great absolute values of ethylene and acetylene pyrolysis velocities on the surface of tungsten carbides in comparison with methane. Criterion of carburizing capability of hydrocarbous relatively to a metal is given on the basis of the results obtained

Irradiation performance of HTGR Biso fertile particles in HFIR experiments HT-17, -18, and -19

International Nuclear Information System (INIS)

Long, E.L. Jr.; Beatty, R.L.; Robbins, J.M.; Kania, M.J.; Eatherly, W.P.

1978-11-01

A series of Biso-coated fertile particles was irradiated in the target facility of the High-Flux Isotope Reactor. The primary objectives of this experiment were to relate the fast-neutron stability of dense propylene-derived pyrocarbons to preferred orientation and to relate irradiation performance to preirradiation characterization values. Coating characterization included x-ray BAF, optical anisotropy, gaseous permeability, small-angle x-ray scattering, and thickness and density determinations. Other objectives were to test Biso-coated large-diameter ThO 2 kernels and coatings derived from propylene diluted with CO 2 rather than argon. Visual examination of the irradiated particles showed that the majority had failed as a result of fast-neutron damage

In-line monitoring of effluents from HTGR fuel particle preparation processes using a time-of-flight mass spectrometer

International Nuclear Information System (INIS)

Lee, D.A.; Costanzo, D.A.; Stinton, D.P.; Carpenter, J.A.; Rainey, W.T. Jr.; Canada, D.C.; Carter, J.A.

1976-08-01

The carbonization, conversion, and coating processes in the manufacture of HTGR fuel particles have been studied with the use of a time-of-flight mass spectrometer. Non-condensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes which have been monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO 2 to UC 2 , buffer and low temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures

Modification V to the computer code, STRETCH, for predicting coated-particle behavior

International Nuclear Information System (INIS)

Valentine, K.H.

1975-04-01

Several modifications have been made to the stress analysis code, STRETCH, in an attempt to improve agreement between the calculated and observed behavior of pyrocarbon-coated fuel particles during irradiation in a reactor environment. Specific areas of the code that have been modified are the neutron-induced densification model and the neutron-induced creep calculation. Also, the capability for modeling surface temperature variations has been added. HFIR Target experiments HT-12 through HT-15 have been simulated with the modified code, and the neutron-fluence vs particle-failure predictions compare favorably with the experimental results. Listings of the modified FORTRAN IV main source program and additional FORTRAN IV functions are provided along with instructions for supplying the additional input data. (U.S.)

Performance of HTGR fertile particles irradiated in HFIR capsule HT-32

International Nuclear Information System (INIS)

Long, E.L. Jr.; Robbins, J.M.; Tiegs, T.N.; Kania, M.J.

1980-04-01

The HT-32 experiment was an uninstrumented capsule irradiated for four cycles in the target position of the High-Flux Isotope Reactor (HFIR). The experiment was designed to: provide supplemental simulated fuel rods for thermal transport and expansion measurements; test fertile kernels with Al 2 O 3 and SiO 2 additives for improved fission product retention; study the stability and permeability of low-temperature isotropic (LTI) pyrocarbon coatings; test Biso- and Triso-coatings derived in a large (0.24-m-dia) coating furnace with a frit distributor; investigate the performance of particles with an outer layer of SiC both as loose particles and as resin-bonded fuel rods; and evaluate high-density alumina as a potential high-temperature thermometry sheathing material

Reoperations following proximal interphalangeal joint nonconstrained arthroplasties.

Science.gov (United States)

Pritsch, Tamir; Rizzo, Marco

2011-09-01

To retrospectively analyze the reasons for reoperations following primary nonconstrained proximal interphalangeal (PIP) joint arthroplasty and review clinical outcomes in this group of patients with 1 or more reoperations. Between 2001 and 2009, 294 nonconstrained (203 pyrocarbon and 91 metal-plastic) PIP joint replacements were performed in our institution. A total of 76 fingers (59 patients) required reoperation (50 pyrocarbon and 26 metal-plastic). There were 40 women and 19 men with an average age of 51 years (range, 19-83 y). Primary diagnoses included osteoarthritis in 35, posttraumatic arthritis in 24, and inflammatory arthritis in 17 patients. There were 21 index, 27 middle, 18 ring, and 10 small fingers. The average number of reoperations per PIP joint was 1.6 (range, 1-4). A total of 45 joints had 1 reoperation, 19 had 2, 11 had 3, and 1 had 4. Extensor mechanism dysfunction was the most common reason for reoperation; it involved 51 of 76 fingers and was associated with Chamay or tendon-reflecting surgical approaches. Additional etiologies included component loosening in 17, collateral ligament failure in 10, and volar plate contracture in 8 cases. Inflammatory arthritis was associated with collateral ligament failure. Six fingers were eventually amputated, 9 had PIP joint arthrodeses, and 2 had resection arthroplasties. The arthrodesis and amputation rates correlated with the increased number of reoperations per finger. Clinically, most patients had no or mild pain at the most recent follow-up, and the PIP joint range-of-motion was not significantly different from preoperative values. Pain levels improved with longer follow-up. Reoperations following primary nonconstrained PIP joint arthroplasties are common. Extensor mechanism dysfunction was the most common reason for reoperation. The average reoperation rate was 1.6, and arthrodesis and amputation are associated with an increasing number of operations. Overall clinical outcomes demonstrated no

Designing the fiber volume ratio in SiC fiber-reinforced SiC ceramic composites under Hertzian stress

International Nuclear Information System (INIS)

Lee, Kee Sung; Jang, Kyung Soon; Park, Jae Hong; Kim, Tae Woo; Han, In Sub; Woo, Sang Kuk

2011-01-01

Highlights: → Optimum fiber volume ratios in the SiC/SiC composite layers were designed under Hertzian stress. → FEM analysis and spherical indentation experiments were undertaken. → Boron nitride-pyrocarbon double coatings on the SiC fiber were effective. → Fiber volume ratio should be designed against flexural stress. -- Abstract: Finite element method (FEM) analysis and experimental studies are undertaken on the design of the fiber volume ratio in silicon carbide (SiC) fiber-reinforced SiC composites under indentation contact stresses. Boron nitride (BN)/Pyrocarbon (PyC) are selected as the coating materials for the SiC fiber. Various SiC matrix/coating/fiber/coating/matrix structures are modeled by introducing a woven fiber layer in the SiC matrix. Especially, this study attempts to find the optimum fiber volume ratio in SiC fiber-reinforced SiC ceramics under Hertzian stress. The analysis is performed by changing the fiber type, fiber volume ratio, coating material, number of coating layers, and stacking sequence of the coating layers. The variation in the stress for composites in relation to the fiber volume ratio in the contact axial or radial direction is also analyzed. The same structures are fabricated experimentally by a hot process, and the mechanical behaviors regarding the load-displacement are evaluated using the Hertzian indentation method. Various SiC matrix/coating/fiber/coating/matrix structures are fabricated, and mechanical characterization is performed by changing the coating layer, according to the introduction (or omission) of the coating layer, and the number of woven fiber mats. The results show that the damage mode changes from Hertzian stress to flexural stress as the fiber volume ratio increases in composites because of the decreased matrix volume fraction, which intensifies the radial crack damage. The result significantly indicates that the optimum fiber volume ratio in SiC fiber-reinforced SiC ceramics should be designed for

Some calculations of the failure statistics of coated fuel particles

International Nuclear Information System (INIS)

Martin, D.G.; Hobbs, J.E.

1977-03-01

Statistical variations of coated fuel particle parameters were considered in stress model calculations and the resulting particle failure fraction versus burn-up evaluated. Variations in the following parameters were considered simultaneously: kernel diameter and porosity, thickness of the buffer, seal, silicon carbide and inner and outer pyrocarbon layers, which were all assumed to be normally distributed, and the silicon carbide fracture stress which was assumed to follow a Weibull distribution. Two methods, based respectively on random sampling and convolution of the variations were employed and applied to particles manufactured by Dragon Project and RFL Springfields. Convolution calculations proved the more satisfactory. In the present calculations variations in the silicon carbide fracture stress caused the greatest spread in burn-up for a given change in failure fraction; kernel porosity is the next most important parameter. (author)

Particle fuel bed tests

International Nuclear Information System (INIS)

Horn, F.L.; Powell, J.R.; Savino, J.M.

1985-01-01

Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Battaglia, Francine

2008-01-01

The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

Interaction between UO2 kernel and pyrocarbon coating in irradiated and unirradiated HTR fuel particles

International Nuclear Information System (INIS)

Drago, A.; Klersy, R.; Simoni, O.; Schrader, K.H.

1975-08-01

Experimental observations on unidirectional UO 2 kernel migration in TRISO type coated particle fuels are reported. An analysis of the experimental results on the basis of data and models from the literature is reported. The stoichiometric composition of the kernel is considered the main parameter that, associated with a temperature gradient, controls the unidirectional kernel migration

Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

International Nuclear Information System (INIS)

Foerthmann, R.; Groos, E.; Gruebmeier, H.

1975-08-01

Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl 2 Si 2 O 8 , BaAl 2 Si 2 O 8 and CsAlSi 2 O 6 in the additional aluminasilica phase of the kernel. (orig.) [de

Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

Energy Technology Data Exchange (ETDEWEB)

Foerthmann, R.; Groos, E.; Gruebmeier, H.

1975-08-15

Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl₂Si₂O₈, BaAl₂Si₂O₈and CsAlSi₂O₆ in the additional alumina-silica phase of the kernel.

In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

Science.gov (United States)

Terrani, K. A.; Silva, C. M.; Kiggans, J. O.; Cai, Z.; Shin, D.; Snead, L. L.

2013-06-01

The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels were examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC-Zr interaction was the least substantial, while the PyC-Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantly faster from what is predicted using literature carbon diffusivity data in ZrC. SiC-Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface.

Microprobe study of fission product behavior in high-burnup HTR fuels

International Nuclear Information System (INIS)

Kleykamp, H.

Electron microprobe analysis of irradiated coated particles with high burnup (greater than 50 percent fima) gives detailed information on the chemical state and the transport behavior of the fission products in UO 2 and UC 2 kernels and in the coatings. In oxide fuel kernels, metallic inclusions and ceramic precipitations are observed. The solubility behavior of the fission products in the fuel matrix has been investigated. Fission product inclusions could not be detected in carbide fuel kernels; post irradiation annealed UC 2 kernels, however, give information on the element combinations of some fission product phases. Corresponding to the chemical state in the kernel, Cs, Sr, Ba, Pd, Te and the rare earths are released easily and diffuse through the entire pyrocarbon coating. These fission products can be retained by a silicon carbide layer. The initial stage of a corrosive attack of the SiC coating by the fission products is evidenced

Tests of candidate materials for particle bed reactors

International Nuclear Information System (INIS)

Horn, F.L.; Powell, J.R.; Wales, D.

1987-01-01

Rhenium metal hot frits and zirconium carbide-coated fuel particles appear suitable for use in flowing hydrogen to at least 2000 K, based on previous tests. Recent tests on alternate candidate cooled particle and frit materials are described. Silicon carbide-coated particles began to react with rhenium frit material at 1600 K, forming a molten silicide at 2000 K. Silicon carbide was extensively attacked by hydrogen at 2066 K for 30 minutes, losing 3.25% of its weight. Vitrous carbon was also rapidly attacked by hydrogen at 2123 K, losing 10% of its weight in two minutes. Long term material tests on candidate materials for closed cycle helium cooled particle bed fuel elements are also described. Surface imperfections were found on the surface of pyrocarbon-coated fuel particles after ninety days exposure to flowing (∼500 ppM) impure helium at 1143 K. The imperfections were superficial and did not affect particle strength

Microscopic search for the carrier phase Q of the trapped planetary noble gases in Allende, Leoville, and Vigarano

Science.gov (United States)

Vis, R. D.; Mrowiec, A.; Kooyman, P. J.; Matsubara, K.; Heymann, D.

2002-10-01

High-resolution transmission electron microscopy micrographs of acid-resistant residues of the Allende, Leoville, and Vigarano meteorites show a great variety of carbon structures: curved and frequently twisted and intertwined graphene sheets, abundant carbon black-like particles, and hollow "sacs". It is suggested that perhaps all of these are carriers for the planetary Q-noble gases in these meteorites. Most of these materials are pyrocarbons that probably formed by the pyrolysis of hydrocarbons either in a gas phase, or on hot surfaces of minerals. An attempt was made to analyze for argon with particle-induced x-ray emission in 143 spots of grains of floating and suspended matter from freeze-dry cycles of an Allende bulk sample in water, and floating "black balls" from sonication in water of samples from the Allende meteorite. The chemical compositions of these particles were obtained, but x-ray signals at the wavelength of argon were obtained on only a few spots.

Microstructure changes and properties of TiC-coated carbon fiber-reinforced carbon composites

International Nuclear Information System (INIS)

Wang Kunjie; Guo Quangui; Zhang Guobing; Shi Jingli; Zhang Hua; Liu Lang

2008-01-01

In the present paper, X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) were used to study distortion of TiC crystals after thermal cycles in plasma environment. Scanning electron microscopy (SEM) was used to observe morphology changes and penetrating cracks in TiC/C coatings. To avoid the cracks and enhance properties of coated carbon fiber-reinforced carbon (C/C) composites, TiC/C composites were prepared as buffer layer to relieve thermal stresses. Thermal cycles indicated that the buffer layer could effectively improve thermal shock resistance of pure TiC coated C/C composites. To study the reason, transmission electron microscopy (TEM) results suggested that TiC particles were uniformly imbedded in pyrocarbon in the buffer layer, which was advantageous to relieve mismatch of coefficient of thermal expansion (CTE) between pure TiC and C/C. Moreover, thermal conductivity tests showed that the buffer layer was in favor of transferring heat loading

Calculations of IAEA-CRP-6 Benchmark Case 1 through 7 for a TRISO-Coated Fuel Particle

International Nuclear Information System (INIS)

Kim, Young Min; Lee, Y. W.; Chang, J. H.

2005-01-01

IAEA-CRP-6 is a coordinated research program of IAEA on Advances in HTGR fuel technology. The CRP examines aspects of HTGR fuel technology, ranging from design and fabrication to characterization, irradiation testing, performance modeling, as well as licensing and quality control issues. The benchmark section of the program treats simple analytical cases, pyrocarbon layer behavior, single TRISO-coated fuel particle behavior, and benchmark calculations of some irradiation experiments performed and planned. There are totally seventeen benchmark cases in the program. Member countries are participating in the benchmark calculations of the CRP with their own developed fuel performance analysis computer codes. Korea is also taking part in the benchmark calculations using a fuel performance analysis code, COPA (COated PArticle), which is being developed in Korea Atomic Energy Research Institute. The study shows the calculational results of IAEACRP- 6 benchmark cases 1 through 7 which describe the structural behaviors for a single fuel particle

Oxidation Kinetics and Strength Degradation of Carbon Fibers in a Cracked Ceramic Matrix Composite

Science.gov (United States)

Halbig, Michael C.

2003-01-01

Experimental results and oxidation modeling will be presented to discuss carbon fiber susceptibility to oxidation, the oxidation kinetics regimes and composite strength degradation and failure due to oxidation. Thermogravimetric Analysis (TGA) was used to study the oxidation rates of carbon fiber and of a pyro-carbon interphase. The analysis was used to separately obtain activation energies for the carbon constituents within a C/SiC composite. TGA was also conducted on C/SiC composite material to study carbon oxidation and crack closure as a function of temperature. In order to more closely match applications conditions C/SiC tensile coupons were also tested under stressed oxidation conditions. The stressed oxidation tests show that C/SiC is much more susceptible to oxidation when the material is under an applied load where the cracks are open and allow for oxygen ingress. The results help correlate carbon oxidation with composite strength reduction and failure.

Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

Energy Technology Data Exchange (ETDEWEB)

Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-02-01

Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).

First Spectroscopic Identification of Pyrocarbonate for High CO2 Flux Membranes Containing Highly Interconnected Three Dimensional Ionic Channels

Science.gov (United States)

2013-01-01

Koura, S. Kohara , K. Takeuchi, S. Takahashi, L. A. Curtiss, M. Grimsditch and M.-L. Saboungi, J. Mol. Struct., 1996, 382, 163–169. 49 L.-J. Chen, X...Cheng, C.-J. Lin and C.-M. Huang, Electrochim. Acta, 2002, 47, 1475–1480. 50 S. Kohara , N. Koura, Y. Idemoto, S. Takahashi, M.-L. Saboungi and L. A

Evaluation of an interlaboratory comparison of the chemical assay of U, Th, oxide coated particles

International Nuclear Information System (INIS)

Tamberg, T.; Thiele, D.; Brodda, B.G.

1981-09-01

The prototype reactor THTR in Schmehausen (Germany, F.R.) burns a (Th,U)O 2 nuclear fuel using 93% enriched uranium. This material is particularly Safeguards sensitive. It was therefore desirable for the Safeguards Analytical Laboratory (SAL) and other laboratories of the Agency Network to collect experience and test their performance in the analysis of such materials. Support was requested from the ''Joint Programme between the IAEA and the Federal Republic of Germany for the Development of Safeguards Techniques'' to perform, as a first step, an interlaboratory comparison of the chemical assay of U and Th in pyrocarbon-coated BISO-type fuel particles. Such an intercomparison was organized under the auspices of the Institut fuer Chemische Technologie (ICT) of the Kernforschungsanlage Juelich GmbH (KFA). SAL prepared a statistical evaluation of the results which was discussed in Vienna in June 1980. The objective of the project was to define the state of the art in the chemical assay of U-Th fuels and the analytical requirements for the sampling of materials of major interest to Agency Safeguards at present

Fission product release profiles from spherical HTR fuel elements at accident temperatures

International Nuclear Information System (INIS)

Schenk, W.; Pitzer, D.; Nabielek, H.

1986-10-01

A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-2500 0 C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO 2 samples at 1400-1800 0 C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-1600 0 C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 1600 0 C are solely due to various forms of contamination; at 1700-1800 0 C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 2000 0 C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 1600 0 C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP) [de

Aminopeptidase Activity from Germinated Jojoba Cotyledons 1

Science.gov (United States)

Johnson, Russell; Storey, Richard

1985-01-01

One major and two minor aminopeptidase activities from germinated jojoba (Simmondsia chinensis) cotyledon extracts were separated by ammonium sulfate precipitation and chromatofocusing. None of the activities were inhibited by 1,10 phenanthroline. The major aminopeptidase, purified 260-fold, showed a pH optimum of 6.9 with leucine-p-nitroanilide as substrate, a molecular weight estimated at 14,200 by electrophoretic analysis, and an isoelectric point of 4.5 according to the chromatofocusing pattern. Activity was inhibited by p-chloromercuribenzoate, slightly stimulated by 1,10 phenanthroline and 2-mercaptoethanol, and not influenced by Mg2+ or diethyl pyrocarbonate. Inhibition by p-chloromercuribenzoate was prevented by the presence of cysteine in the assay. Leucine-p-nitroanilide and leucine-β-naphthylamide were the most rapidly hydrolyzed of 11 carboxy-terminal end blocked synthetic substrates tested. No activity on endopeptidase or carboxypeptidase specific substrates was detected. The major aminopeptidase showed activity on a saline soluble, jojoba seed protein preparation and we suggest a possible physiological role for the enzyme in the concerted degradation of globulin reserve proteins during cotyledon senescence. PMID:16664465

Aminopeptidase activity from germinated jojoba cotyledons.

Science.gov (United States)

Johnson, R; Storey, R

1985-11-01

One major and two minor aminopeptidase activities from germinated jojoba (Simmondsia chinensis) cotyledon extracts were separated by ammonium sulfate precipitation and chromatofocusing. None of the activities were inhibited by 1,10 phenanthroline.The major aminopeptidase, purified 260-fold, showed a pH optimum of 6.9 with leucine-p-nitroanilide as substrate, a molecular weight estimated at 14,200 by electrophoretic analysis, and an isoelectric point of 4.5 according to the chromatofocusing pattern. Activity was inhibited by p-chloromercuribenzoate, slightly stimulated by 1,10 phenanthroline and 2-mercaptoethanol, and not influenced by Mg(2+) or diethyl pyrocarbonate. Inhibition by p-chloromercuribenzoate was prevented by the presence of cysteine in the assay. Leucine-p-nitroanilide and leucine-beta-naphthylamide were the most rapidly hydrolyzed of 11 carboxy-terminal end blocked synthetic substrates tested. No activity on endopeptidase or carboxypeptidase specific substrates was detected. The major aminopeptidase showed activity on a saline soluble, jojoba seed protein preparation and we suggest a possible physiological role for the enzyme in the concerted degradation of globulin reserve proteins during cotyledon senescence.

Essential histidyl residues at the active site(s) of sucrose-phosphate synthase from Prosopis juliflora.

Science.gov (United States)

Sinha, A K; Pathre, U V; Sane, P V

1998-11-10

Chemical modification of sucrose-phosphate synthase (EC 2.4.1.14) from Prosopis juliflora by diethyl pyrocarbonate (DEP) and photo-oxidation in the presence of rose bengal (RB) which modify the histidyl residues of the protein resulted in the inactivation of the enzyme activity. This inactivation was dependent on the concentration of the modifying reagent and the time of incubation and followed pseudo-first order kinetics. For both the reagents, the inactivation was maximum at pH 7.5, which is consistent with the involvement and presence of histidine residues at the active site of the enzyme. Substrates, UDPG and F6P protected the enzyme against the inactivation by the modifying reagents suggesting that the histidine residues may be involved in the binding of these substrates and are essential for the catalytic activity. Specificity of DEP was indicated by an increase in absorbance at 240 nm along with concomitant inactivation of the enzyme and reactivation of the modified enzyme by hydroxylamine. These results strongly suggest the presence of histidine residue(s) at or near the active site of the enzyme.

DC electrical conductivity of silicon carbide ceramics and composites for flow channel insert applications

International Nuclear Information System (INIS)

Katoh, Y.; Kondo, S.; Snead, L.L.

2009-01-01

High purity chemically vapor-deposited silicon carbide (SiC) and 2D continuous SiC fiber, chemically vapor-infiltrated SiC matrix composites with pyrocarbon interphases were examined. Specifically, temperature dependent (RT to 800 deg. C) electrical conductivity and the influence of neutron irradiation were measured. The influence of neutron irradiation on electrical properties appeared very strong for the SiC of this study, typically resulting in orders lower ambient conductivity and steeper temperature dependency of this conductivity. For the 2D composites, through-thickness (normal to the fiber axis') electrical conductivity was dominated by bypass conduction via interphase network at relatively low temperatures, whereas conduction through SiC constituents dominated at higher temperatures. Through-thickness electrical conductivity of neutron-irradiated 2D SiC composites with thin PyC interphase, currently envisioned for flow channel insert application, will likely in the order of 10 S/m at the appropriate operating temperature. Mechanisms of electrical conduction in the composites and irradiation-induced modification of electrical conductivity of the composites and their constituents are discussed.

Preparation and physical properties of vapour-deposited carbon-carbon composites

International Nuclear Information System (INIS)

Loll, Philippe

1976-01-01

In its first part, this research thesis reports a bibliographical study on methods of preparation of various types of vapour-deposited (CVD) carbons, and the author notices that only structure and texture properties of these macroscopically homogeneous pyro-carbons have been studied in detail. For a better understanding of the behaviour of carbon-carbon composites, this thesis thus reports the study of the relationships between physical properties, macroscopic texture and microscopic structure. A densification installation and methods of characterisation have been developed. The fabrication process and its installation are presented (oven with its temperature and gas rate controls, study of its thermal gradient, substrate, heat treatments), and the study and characterisation of carbon-carbon composites are reported: structure and texture properties (studied by optic and scanning electronic microscopy, density measurements, and X-ray diffraction), physical properties (electronic paramagnetic resonance, static magnetism, electric and thermal conductivity). In the last part, the author comments and discusses the obtained results: conditions of preparation, existence, physical properties of the different observed microstructures [fr

Calcination, Reduction and Sintering of ADU Spheres for HTGR Fuel

International Nuclear Information System (INIS)

Jeong, Kyung Chai; Eom, Sung Ho; Kim, Yeon Ku; Kim, Woong Ki; Kim, Young Min; Lee, Young Woo; Kim, Ju Hee; Cho, Hyo Jin; Cho, Moon Seoung

2011-01-01

The international oil market is again in turmoil in accordance with the increasing of human needs and energy consumption. Soaring oil prices, fears of supply security, and climate change are concerned becoming more concrete make for an uncertain energy future. In this view point, nuclear energy is an important, yet controversial option for energy supply. High Temperature Gas Reactor will play a dominant role in the worldwide fleet of nuclear reactors of the next decade for electricity production and high temperature heat. HTGR have two reactor types which use the basic fuel concept based on the dispersion of TRISO coated particles in graphite in shown Fig.1. The TRISO coated particle for these purposes is prepared with pyro-carbon and silicone carbide coatings on a spherical UO 2 kernel surface as fissile material. The TRISO fuel particle consists of a microsphere (i.e., UO 2 kernel) of nuclear material: encapsulated by multiple layers of pyro-carbon and a SiC layer. This multiple coating layers system has been engineered to retain the fission products generated by fission of the nuclear material in the kernel during normal operation and all licensing basis events over the design lifetime of the fuel. UO 2 kernels are produced by using the modified sol-gel process, a wet process, generally known as the GSP method. Wet chemical processes are flexible in producing kernels of different size and chemical composition with high throughout and yield, good spherical shape, and narrow size distribution. This chemical processing route is well-known to the potential kernel fabrication processes. The principle, as set out in Fig.2, involves first of all preparing a pseudo-sol(also known as a 'broth') from an initial uranyl nitrate solution . This broth solution is obtained through addition of a number of additives, as determined by process know-how, including a soluble organic polymer, that are subsequently gels into droplets and are dispersed for ADU precipitation. The

Acceptance Test Data for BWXT Coated Particle Batches 93172B and 93173B—Defective IPyC and Pyrocarbon Anisotropy

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Coated particle batches J52O-16-93172B and J52O-16-93173B were produced by Babcock and Wilcox Technologies (BWXT) as part of the production campaign for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), but were not used in the final fuel composite. However, these batches may be used as demonstration production-scale coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93172A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017b]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93172B).

Review of fuel element development for nuclear rocket engines

International Nuclear Information System (INIS)

Taub, J.M.

1975-06-01

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program involving uranium-loaded graphite fuels, hydrogen propellant, and a target exhaust temperature of approximately 2500 0 C. A very extensive uranium-loaded graphite fuel element technology evolved from the program. Selection and composition of raw materials for the extrusion mix had to be coupled with heat treatment studies to give optimum element properties. The highly enriched uranium in the element was incorporated as UO 2 , pyrocarbon-coated UC 2 , or solid solution UC . ZrC particles. An extensive development program resulted in successful NbC or ZrC coatings on elements to withstand hydrogen corrosion at elevated temperatures. Hot gas, thermal shock, thermal stress, and NDT evaluation procedures were developed to monitor progress in preparation of elements with optimum properties. Final evaluation was made in reactor tests at NRDS. Aerojet-General, Westinghouse Astronuclear Laboratory, and the Oak Ridge Y-12 Plant of Union Carbide Nuclear Company entered the program in the early 1960's, and their activities paralleled those of LASL in fuel element development. (U.S.)

Involvement of Histidine Residue His382 in pH Regulation of MCT4 Activity.

Directory of Open Access Journals (Sweden)

Shotaro Sasaki

Full Text Available Monocarboxylate transporter 4 (MCT4 is a pH-dependent bi-directional lactate transporter. Transport of lactate via MCT4 is increased by extracellular acidification. We investigated the critical histidine residue involved in pH regulation of MCT4 function. Transport of lactate via MCT4 was measured by using a Xenopus laevis oocyte expression system. MCT4-mediated lactate transport was inhibited by Zn2+ in a pH physiological condition but not in an acidic condition. The histidine modifier DEPC (diethyl pyrocarbonate reduced MCT4 activity but did not completely inactivate MCT4. After treatment with DEPC, pH regulation of MCT4 function was completely knocked out. Inhibitory effects of DEPC were reversed by hydroxylamine and suppressed in the presence of excess lactate and Zn2+. Therefore, we performed an experiment in which the extracellular histidine residue was replaced with alanine. Consequently, the pH regulation of MCT4-H382A function was also knocked out. Our findings demonstrate that the histidine residue His382 in the extracellular loop of the transporter is essential for pH regulation of MCT4-mediated substrate transport activity.

Optimization of phenol extraction procedures for preparation of RNA from mammalian lymphoid organs

Energy Technology Data Exchange (ETDEWEB)

Griffin, G.D.; Sellin, H.G.; Novelli, G.D.

1978-07-01

Methods have been developed to optimize the extraction of RNA from mammalian lymphoid organs (spleen) with respect to both quantity and quality of RNA and with minimal DNA contamination. Nuclease inhibitors, including diethyl pyrocarbonate, polyvinyl sulfate, and bentonite were used in the initial disruption of the tissue, which was accomplished by blender, Dounce homogenizer, or preparation of a cell suspension. Seven buffer systems, varying with respect to pH, detergent, and NaCl concentration, were used in the initial extraction with phenol, and the temperature of extraction was also varied. Protocols involving the selective use of naphthalene 1.5-disulfonic acid and sodium dodecyl sulfate were developed to provide an initial RNA extract with minimal DNA content. Dounce homogenization, followed by separate treatment of nuclear and cytosol fractions, was found to be the most effective technique, both in terms of RNA yield (averaging 76%) and the quality of RNA recovered (as judged by gel electrophoresis). RNA from blender preparations contained larger amounts of DNA and RNA yield was decreased to 54%. RNA extracted from spleen cell suspensions was of poor quality and gave very poor yield (27%).

Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

International Nuclear Information System (INIS)

Nickel, H.

1985-08-01

Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

Energy Technology Data Exchange (ETDEWEB)

Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-05-01

Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).

Fuel development for reactors of new generation in Ukraine

International Nuclear Information System (INIS)

Odeychuk, N.P.

2006-01-01

elements development with fuel on a basis: Metal: uranium, alloys of uranium; Ceramic: uranium dioxide, thorium dioxide, uranium carbonitride, uranium oxycarbide, mixed oxide of uranium and thorium. The special attention is given to discussion of the basic technological schemes of reception of the fuel microspheres, coated particles and spherical fuel elements for HTGR. Features of reception carbongraphite materials and products by the methods of volumetric gas-phase condensation of porous preparations by pyrocarbon are considered. Results of investigations of the basic fuel elements characteristics and their components, materials and products with pyrocarbon binding, including in conditions of reactor irradiations are discussed. The review concerning the experience of the development the fuel elements with fuel based on metal uranium is given. In NSC KIPT constructions and manufacturing techniques of components for active zones of new perspective directions of atomic engineering are created and proved, also was laid the foundation for the base design and technological decisions for the fourth generation nuclear reactors

Data Compilation for AGR-1 Baseline Coated Particle Composite LEU01-46T

International Nuclear Information System (INIS)

Hunn, John D.; Lowden, Richard Andrew

2006-01-01

This document is a compilation of characterization data for the AGR-1 baseline coated particle composite LEU01-46T, a composite of four batches of TRISO-coated 350 (micro)m 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with a ∼ 50% dense carbon buffer layer (100 (micro)m nominal thickness) followed by a dense inner pyrocarbonlayer (40 (micro)m nominal thickness) followed by a SiC layer (35 (micro)m nominal thickness) followed by another dense outer pyrocarbon layer (40 (micro)m nominal thickness). The coated particles, were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program to be put into compacts for insertion in the first irradiation test capsule, AGR-1. The kernels were obtained from BWXT and identified as composite (G73D-20-69302). The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). Additional particle batches were coated with only buffer or buffer plus inner pyrocarbon (IPyC) layers using similar process conditions as used for the full TRISO batches comprising the LEU01-46T composite. These batches were fabricated in order to qualify that the process conditions used for buffer and IPyC would produce acceptable densities, as described in sections 8 and 9. These qualifying batches used 350 (micro)m natural uranium oxide/uranium carbide kernels (NUCO). The kernels were obtained from BWXT and identified as composite G73B-NU-69300. The use of NUCO surrogate kernels is not expected to significantly effect the densities of the buffer and IPyC coatings. Confirmatory batches using LEUCO kernels from G73D-20-69302 were coated and characterized to verify this assumption. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380, Rev. 6) provides the requirements necessary for acceptance

Radiation response of SiC-based fibers

Energy Technology Data Exchange (ETDEWEB)

Youngblood, G.E.; Jones, R.H. [Battelle Pacific Northwest Labs., Richland, WA (United States); Kohyama, A. [Inst. of Advanced Energy, Kyoto Univ. (Japan); Snead, L.L. [Oak Ridge National Lab., TN (United States)

1998-10-01

Loss of strength in irradiated fiber-reinforced SiC/SiC composite generally is related to degradation in the reinforcing fiber. To assess fiber degradation, the density and length changes were determined for four types of SiC-based fibers (Tyranno, Nicalon CG, Hi Nicalon and Dow X) after high temperature (up to 1000 C) and high dose (up to 80 dpa-SiC) irradiations. For the fibers with nonstoichiometric compositions (the first three types in the list), the fiber densities increased from 6% to 12%. In contrast, a slight decrease in density (<1%) was observed for the Dow X fiber with a quasi-stoichiometric composition. Fiber length changes (0-5.6% shrinkage) suggested small mass losses (1-6%) had occurred for irradiated uncoated fibers. In contrast, excessive linear shrinkage of the pyrocarbon-coated Nicalon CG and Tyranno fibers (7-9% and 16-32%, respectively) indicated that much larger mass losses (11-84%) had occurred for these coated fibers. Crystallization and crystal growth were observed to have taken place at fiber surfaces by SEM and in the bulk by XRD, moreso for irradiated Nicalon CG than for Hi Nicalon fiber. The radiation response of the quasi-stoichiometric Dow X fiber was the most promising. Further testing of this type fiber is recommended. (orig.) 11 refs.

Radioisotope space power generator annual report, July 1, 1974--June 30, 1975

International Nuclear Information System (INIS)

Elsner, N.B.; Chin, J.; Staley, H.G.

1976-01-01

The Isotec Technology Program for FY-75 concentrated on materials development efforts in two areas: TPM-217 P-type material and SiGe technology. TPM-217 P-type material is a 3M proprietary thermoelectric material whose principal components are Cu, Ag, and Se. The usefulness of TPM-217 P-type selenide in thermoelectric converters depends on its dimensional, electrical, and thermal stability at high temperature and its compatibility with other converter component materials in a low-pressure environment. Experimental efforts were directed toward determining (1) the vapor species above TPM-217 from 700 0 to 1100 0 C, (2) the weight loss rate for TPM-217 at 900 0 C in vacuo, and (3) the stability of TPM-217 material in contact with Mo, Fe, 316 stainless steel, and pyrocarbon. The Si-Ge program is a continuation of the experimental work performed during FY-74. The development of coatings to suppress the vaporization of SiMo and SiGe continued. Techniques for applying ion-plated and chemical-vapor-deposited coatings of Si 3 N 4 and (Si,Al)N alloys on SiMo were examined. Methods of controlling morphology and the chemical composition of these coatings were developed. Rates of vaporization for coated samples at 1100 0 C were measured

Performance assessment of the (Th,U)O2 HTI-Biso coated particle under PNP/HHT irradiation conditions

International Nuclear Information System (INIS)

Kania, M.J.; Nickel, H.

1980-11-01

The HTI Biso Particle, Variant-I: consisting of a dense 400 μm-diameter (Th,U)O 2 -kernel with a Biso coating using a methane derived pyrocarbon layer (HTI), is a candidate fuel for the advanced PNP/HHT High Temperature Reactor systems. This report presents the results of a comprehensive performance assessment of Variant-I represented by six relevant particle batches irradiated in 12 accelerated irradiation experiments. Fuel performance was judged based upon PNP/HHT qualification requirements with regard to in-reactor operating conditions and end-of-life (EOL) coated particle failure fraction. Fuel operating conditions in each irradiation experiment were obtained from two sources: 1) a thorough review of all available irradiation data on each experiment; and 2) a two-dimensional (R,theta) thermal modeling computer code, R2KTMP, was developed to calculate fuel operating temperature distributions within spherical elements. End-of-life particle failure fractions were determined from: gaseous fission product release, based on in-reactor R/B measurements and postirradiation annealing and room temperature investigations; solid fission product release, from single particle 137 Cs release into fuel element matrix and hot-gaseous chlorine leaching; and visual and ceramographic examinations. Failure fractions determined by solid fission product release yielded values 2-35 times higher than those determined by gaseous fission product release. (orig.) [de

Irradiation performance of HTGR fertile fuel in HFIR target capsules HT-12 through HT-15. Part I. Experiment description and fission product behavior

International Nuclear Information System (INIS)

Kania, M.J.; Lindemer, T.B.; Morgan, M.T.; Robbins, J.M.

1977-02-01

Sixteen types of Biso-coated designs, on ThO 2 kernels, were irradiated in High Flux Isotope Reactor target capsules HT-12 through HT-15. The report addresses the description of the experiment and extensive postirradiation analyses and experiments to determine fertile-particle burnup, fuel coating failures, and fission product behavior. Several low-temperature isotropic (LTI) pyrocarbon coatings, which ''survived'' according to visual inspection, were shown to have developed permeability during irradiation. These particles were irradiated at temperatures approximately equal to 1250 0 C and to burnups equal to or greater than 8 percent fission per initial heavy-metal atom (FIMA). No evidence of permeability was found in similar particles irradiated at temperatures approximately equal to 1550 0 C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000 0 C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300 0 C in excess of design limits of 900 0 C (low-temperature magazines) and 1250 0 C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10 22 neutrons/cm 2 produced severe degradation and swelling of the Poco graphite magazines and sample holders

Fluidized bed deposition and evaluation of silicon carbide coatings on microspheres

International Nuclear Information System (INIS)

Federer, J.I.

1977-01-01

The fuel element for the HTGR is an array of closely packed fuel microspheres in a carbonaceous matrix. A coating of dense silicon carbide (SiC), along with pyrocarbon layers, is deposited on the fueled microspheres to serve as a barrier against diffusion of fission products. The microspheres are coated with silicon carbide in a fluidized bed by reaction of methyltrichlorosilane (CH 3 SiCl 3 or MTS) and hydrogen at elevated temperatures. The principal variables of coating temperature and reactant gas composition (H 2 /MTS ratio) have been correlated with coating rate, morphology, stoichiometry, microstructure, and density. The optimum temperature for depositing highly dense coatings is in the range 1475 to 1675 0 C. Lower temperatures result in silicon-rich deposits, while higher temperatures may cause unacceptable porosity. The optimum H 2 /MTS ratio for highly dense coatings is 20 or more (approximately 5% MTS or less). The amount of grown-in porosity increases as the H 2 /MTS ratio decreases below 20. The requirement that the H 2 /MTS ratio be about 20 or more imposes a practical restraint on coating rate, since increasing the total flow rate would eventually expel microspheres from the coating tube. Evaluation of stoichiometry, morphology, and microstructure support the above mentioned optimum conditions of temperature and reactant gas composition. 18 figures, 3 tables

Interaction between uranium oxide alloyed with Al2O3·SiO2 and pyrocarbon coating during irradiation of micro fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Khromov, Y.F.; Svistunov, D.E.; Chuiko, E.E.

1989-01-01

The thermodynamics of the interaction between uranium oxide and carbon was previously studied in the presence of Al 2 O 3 ·SiO 2 , SiC, and UC 1.86 ; in this case, the quantity of the reacting substances does not have any effect on the attainment of the equilibrium state. Based on the obtained results, it is interesting to study the characteristic features of the interaction between the alloyed UO x cores (kernels) with the PyC-coating under the conditions involving irradiation of the micro fuel elements with thermal neutrons and the formation of solid fission products. The data concerning the characteristics of a micro fuel element (the weight of the core, its composition, etc.) are useful for carrying out a quantitative evaluation of the additives required for fixing the alkali-earth fission products by obtaining stable compounds of aluminosilicates with Ba, Sr, Rb, and Cs at different levels of depletion (burnup) of the oxide fuel. An analysis of the interaction processes in such a complex system as the irradiated alloyed uranium oxide fuel located in a micro fuel element is carried out by comparing the chemical potential of oxygen (RT ln P O 2 ) for the competing constituents of the system

Production and Characterization of Organic Solvent-Tolerant Cellulase from Bacillus amyloliquefaciens AK9 Isolated from Hot Spring.

Science.gov (United States)

Irfan, Muhammad; Tayyab, Ammara; Hasan, Fariha; Khan, Samiullah; Badshah, Malik; Shah, Aamer Ali

2017-08-01

A cellulase-producing bacterium, designated as strain AK9, was isolated from a hot spring of Tatta Pani, Azad Kashmir, Pakistan. The bacterium was identified as Bacillus amyloliquefaciens through 16S rRNA sequencing. Cellulase from strain AK9 was able to liberate glucose from soluble cellulose and carboxymethyl cellulose (CMC). Enzyme was purified through size exclusion chromatography and a single band of ∼47 kDa was observed on sodium dodecyl sulfate-polyacrylamide gel electrophoresis (SDS-PAGE). The enzyme was purified with recovery of 35.5%, 3.6-fold purity with specific activity of 31 U mg -1 . The purified cellulase retained its activity over a wide range of temperature (50-70 °C) and pH (3-7) with maximum stability at 60 °C and pH 5.0. The activity inhibited by ethylenediaminetetraacetic acid (EDTA), suggested that it was metalloenzyme. Diethyl pyrocarbonate (DEPC) and β-mercaptoethanol significantly inhibited cellulase activity that revealed the essentiality of histidine residues and disulfide bonds for its catalytic function. It was stable in non-ionic surfactants, in the presence of various metal ions, and in water-insoluble organic solvents. Approximately 9.1% of reducing sugar was released after enzymatic saccharification of DAP-pretreated agro-residue, compared to a very low percentage by autohydrolysis treatment. Hence, it is concluded that cellulase from B. amyloliquefaciens AK9 can potentially be used in bioconversion of lignocellulosic biomass to fermentable sugars.

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

Energy Technology Data Exchange (ETDEWEB)

Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-03-01

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).

Preparation and properties of carbonaceous products prepared by the cracking of natural gas; Fabrication et proprietes de corps carbones prepares par craquage de gaz naturel

Energy Technology Data Exchange (ETDEWEB)

Blum, P L; Bochirol, L; Rappeneau, J; Cornuault, P; Blanchard, R; Moreau, C [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

Complete results are presented of tests recently carried out at the Grenoble Nuclear Research Centre in an attempt to transform natural gas (gas from Lacq), used as a source of pyrolytic carbon, into carbonaceous products with improved properties. Several methods have been studied: 1 - By using normal grade industrial graphite as support (density of about 1.50) products having densities of about 1.80 are obtained. Their open porosity (6 to 7 per cent) is lower than that of conventional graphites, and several of their characteristics are more or less equal to, or better than those obtained from a double impregnation witch pitch. 2 - The use as supporting material of the semi-products ('cooked') usually used for graphite production does not lead to satisfactory results. The main reasons for this are given. 3 - A new process, called 'BB5', has been developed. The starting materials here are powdered products (petrol coke or graphite) which are put into shape with the use of a binder which can be dispersed in water. The supports thus produced make it possible, because of their porous structure, to make the most of the densification produced by cracking natural gas below 1000 deg. C. The products obtained, which are made up of equal amounts approximately of the supporting material and of pyrocarbon can attain densities of over 1.90. Their very low open porosity can be reduced almost to zero and their impermeability is then excellent (k=10{sup -8} cm{sup 2}sec{sup -1}). They have also a remarkably high resistance to compression, values of 15 to 20 kg/mm{sup 2} being obtained for those carbons which have not undergone a final graphitization treatment. Some examples are given of possible nuclear applications of the materials produced in this manner. (authors) [French] On expose l'ensemble des essais effectues recemment au CEN-G pour preparer, a partir de gaz naturel (gaz de Lacq) comme source de carbone pyrolytique, des corps carbones de caracteristiques ameliorees

PARFUME Theory and Model basis Report

Energy Technology Data Exchange (ETDEWEB)

Darrell L. Knudson; Gregory K Miller; G.K. Miller; D.A. Petti; J.T. Maki; D.L. Knudson

2009-09-01

The success of gas reactors depends upon the safety and quality of the coated particle fuel. The fuel performance modeling code PARFUME simulates the mechanical, thermal and physico-chemical behavior of fuel particles during irradiation. This report documents the theory and material properties behind vari¬ous capabilities of the code, which include: 1) various options for calculating CO production and fission product gas release, 2) an analytical solution for stresses in the coating layers that accounts for irradiation-induced creep and swelling of the pyrocarbon layers, 3) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or a prismatic block core, as well as through the layers of each analyzed particle, 4) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, and kernel migration (or amoeba effect), 5) two independent methods for determining particle failure probabilities, 6) a model for calculating release-to-birth (R/B) ratios of gaseous fission products that accounts for particle failures and uranium contamination in the fuel matrix, and 7) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. The accident condi¬tion entails diffusion of fission products through the particle coating layers and through the fuel matrix to the coolant boundary. This document represents the initial version of the PARFUME Theory and Model Basis Report. More detailed descriptions will be provided in future revisions.

Nanoscale multilayered and porous carbide interphases prepared by pressure-pulsed reactive chemical vapor deposition for ceramic matrix composites

International Nuclear Information System (INIS)

Jacques, S.; Jouanny, I.; Ledain, O.; Maillé, L.; Weisbecker, P.

2013-01-01

In Ceramic Matrix Composites (CMCs) reinforced by continuous fibers, a good toughness is achieved by adding a thin film called “interphase” between the fiber and the brittle matrix, which acts as a mechanical fuse by deflecting the matrix cracks. Pyrocarbon (PyC), with or without carbide sub-layers, is typically the material of choice to fulfill this role. The aim of this work was to study PyC-free nanoscale multilayered carbide coatings as interphases for CMCs. Nanoscale multilayered (SiC–TiC) n interphases were deposited by pressure-Pulsed Chemical Vapor Deposition (P-CVD) on single filament Hi-Nicalon fibers and embedded in a SiC matrix sheath. The thicknesses of the carbide interphase sub-layers could be made as low as a few nanometers as evidenced by scanning and transmission electron microscopy. By using the P-ReactiveCVD method (P-RCVD), in which the TiC growth involves consumption of SiC, it was not only possible to obtain multilayered (SiC–TiC) n films but also TiC films with a porous multilayered microstructure as a result of the Kirkendall effect. The porosity in the TiC sequences was found to be enhanced when some PyC was added to SiC prior to total RCVD consumption. Because the porosity volume fraction was still not high enough, the role of mechanical fuse of the interphases could not be evidenced from the tensile curves, which remained fully linear even when chemical attack of the fiber surface was avoided.

Participatory role of zinc in structural and functional characterization of bioremediase: a unique thermostable microbial silica leaching protein.

Science.gov (United States)

Chowdhury, Trinath; Sarkar, Manas; Chaudhuri, Biswadeep; Chattopadhyay, Brajadulal; Halder, Umesh Chandra

2015-07-01

A unique protein, bioremediase (UniProt Knowledgebase Accession No.: P86277), isolated from a hot spring bacterium BKH1 (GenBank Accession No.: FJ177512), has shown to exhibit silica leaching activity when incorporated to prepare bio-concrete material. Matrix-assisted laser desorption ionization mass spectrometry analysis suggests that bioremediase is 78% homologous to bovine carbonic anhydrase II though it does not exhibit carbonic anhydrase-like activity. Bioinformatics study is performed for understanding the various physical and chemical parameters of the protein which predicts the involvement of zinc encircled by three histidine residues (His94, His96 and His119) at the active site of the protein. Isothermal titration calorimetric-based thermodynamic study on diethyl pyrocarbonate-modified protein recognizes the presence of Zn(2+) in the enzyme moiety. Exothermic to endothermic transition as observed during titration of the protein with Zn(2+) discloses that there are at least two binding sites for zinc within the protein moiety. Addition of Zn(2+) regains the activity of EDTA chelated bioremediase confirming the presence of extra binding site of Zn(2+) in the protein moiety. Revival of folding pattern of completely unfolded urea-treated protein by Zn(2+) explains the participatory role of zinc in structural stability of the protein. Restoration of the λ max in intrinsic fluorescence emission study of the urea-treated protein by Zn(2+) similarly confirms the involvement of Zn in the refolding of the protein. The utility of bioremediase for silica nanoparticles preparation is observed by field emission scanning electron microscopy.

Nanoscale multilayered and porous carbide interphases prepared by pressure-pulsed reactive chemical vapor deposition for ceramic matrix composites

Science.gov (United States)

Jacques, S.; Jouanny, I.; Ledain, O.; Maillé, L.; Weisbecker, P.

2013-06-01

In Ceramic Matrix Composites (CMCs) reinforced by continuous fibers, a good toughness is achieved by adding a thin film called "interphase" between the fiber and the brittle matrix, which acts as a mechanical fuse by deflecting the matrix cracks. Pyrocarbon (PyC), with or without carbide sub-layers, is typically the material of choice to fulfill this role. The aim of this work was to study PyC-free nanoscale multilayered carbide coatings as interphases for CMCs. Nanoscale multilayered (SiC-TiC)n interphases were deposited by pressure-Pulsed Chemical Vapor Deposition (P-CVD) on single filament Hi-Nicalon fibers and embedded in a SiC matrix sheath. The thicknesses of the carbide interphase sub-layers could be made as low as a few nanometers as evidenced by scanning and transmission electron microscopy. By using the P-ReactiveCVD method (P-RCVD), in which the TiC growth involves consumption of SiC, it was not only possible to obtain multilayered (SiC-TiC)n films but also TiC films with a porous multilayered microstructure as a result of the Kirkendall effect. The porosity in the TiC sequences was found to be enhanced when some PyC was added to SiC prior to total RCVD consumption. Because the porosity volume fraction was still not high enough, the role of mechanical fuse of the interphases could not be evidenced from the tensile curves, which remained fully linear even when chemical attack of the fiber surface was avoided.

Effects of Processing Parameters on the Density and Microstructure of Pyrolytic Carbon

International Nuclear Information System (INIS)

Kim, Weon Ju; Park, Jeong Nam; Park, Jong Hoon; Cho, Moon Sung; Lee, Young Woo; Park, Ji Yeon

2007-01-01

Chemical vapor deposition (CVD) of pyrolytic carbon (PyC) and silicon carbide (SiC) has been applied to TRISO-coated fuel particles for high-temperature gas-cooled reactors (HTGR). The porous PyC coating layer, called the buffer layer, attenuates fission recoils and provides void volume for gaseous fission products and carbon monoxide. The inner PyC layer acts as a containment to gaseous products. The outer PyC layer protects the SiC coating layer by inducing a compressive stress and provides chemical compatibility with a graphite matrix in the fuel compact. The PyC layers undergo shrinkage due to neutron irradiation, affecting the design and modeling of fuel particles. Because the dimensional change of PyC depends on the detailed microstructure of PyC, it differs from one fabrication route to another one. This requires a new design of irradiation experiment applicable to spherical objects and leads to an international collaborative work called PYCASSO (PYrocarbon irradiation for Creep And Swelling/Shrinkage of Objects). KAERI proposed four different types of PyC layers coated on ZrO 2 particles, buffer with a density of 1.0 and dense PyCs with densities of 1.7, 1.9 and 2.1 g/cm 3 , for the irradiation experiment. In this study, we fabricated PyC-coated particles with various coating densities for supporting the PYCASSO experiment. We also investigated effects of processing parameters such as temperature, hydrocarbon concentration and gas flow rate on the density and microstructure of the PyC layer

Hydrogen adsorption in the series of carbon nanostructures: Graphenes-graphene nanotubes-nanocrystallites

Science.gov (United States)

Soldatov, A. P.; Kirichenko, A. N.; Tat'yanin, E. V.

2016-07-01

A comparative analysis of hydrogen absorption capability is performed for the first time for three types of carbon nanostructures: graphenes, oriented carbon nanotubes with graphene walls (OCNTGs), and pyrocarbon nanocrystallites (PCNs) synthesized in the pores of TRUMEM ultrafiltration membranes with mean diameters ( D m) of 50 and 90 nm, using methane as the pyrolized gas. The morphology of the carbon nanostructures is studied by means of powder X-ray diffraction, X-ray photoelectron spectroscopy (XPS), Raman spectroscopy, and transmission electron microscopy (TEM). Hydrogen adsorption is investigated via thermogravimetric analysis (TGA) in combination with mass-spectrometry. It is shown that only OCNTGs can adsorb and store hydrogen, the desorption of which under atmospheric pressure occurs at a temperature of around 175°C. Hydrogen adsorption by OCNTGs is quantitatively determined and found to be about 1.5% of their mass. Applying certain assumptions, the relationship between the mass of carbon required for the formation of single-wall OCNTGs in membrane pores and the surface area of pores is established. Numerical factor Ψ = m dep/ m calc, where m dep is the actual mass of carbon deposited upon the formation of OCNTGs and mcalc is the calculated mass of carbon necessary for the formation of OCNTGs is introduced. It is found that the dependence of specific hydrogen adsorption on the magnitude of the factor has a maximum at Ψ = 1.2, and OCNTGs can adsorb and store hydrogen in the interval 0.4 to 0.6 hydrogen adsorption and its relationship to the structure of carbon nanoformations are examined.

Improvements in or relating to the manufacture of compact bodies from particulate material

International Nuclear Information System (INIS)

Lefevre, R.L.R.; Barbier, Y.M.J.; Thomas, J.P.; Sturge, D.W.J.

1976-01-01

It is stated that a problem arises in the manufacture of compact bodies of nuclear fuel from fuel particles. The particles may comprise spheroidal kernels of sintered UO 2 coated with layers of pyrocarbon and SiC and overcoated with layers of soft graphite. These particles are compressed in dies to form compact bodies, the layers of overcoating deforming to fill the spaces between the fuel particles. When the compact bodies are to be of elongated form pressure is applied axially through rams entering the ends of the die cavity, but this unfortunately leads to a variation in volume loading and matrix density of the particles, so that those particles and associated matrix at the ends of the elongated body are compacted to a higher degree than those at the center. Attempts to rectify this by increasing the pressure of compaction often results in breakage of the particles subjected to the greatest pressure. The method of manufacture described seeks to overcome this difficulty. The material is placed in a shaped die cavity, that partly defines the shape of the body, and is compacted between oppositely acting pairs of rams that extend over the full length of the cavity. Forces are then applied normal to the longitudinal axis of the cavity. The rams are shaped to combine with the cavity to define the shape of the compacted body. The addition of some end-compaction is advantageous, and for this purpose additional forces may be applied in the direction of the longitudinal axis. The die cavity may contain a mandrel arranged so that a hollow compact body is formed by compaction. (U.K.)

HTGR fuel rods: carbon-carbon composites designed for high weight and low strength

International Nuclear Information System (INIS)

Bullock, R.E.

1977-01-01

The evolution of the process for fabricating fuel rods for the high-temperature gas-cooled reactor (HTGR) by injection and carbonization of a thermoplastic matrix that bonds close-packed beds of pyrocarbon-coated fuel particles together is reviewed for the fresh-fuel cycle, and a variant process involving a thermosetting matrix that would allow free-standing carbonization of refabricated fuel is discussed. Previous attempts to fabricate such injection-bonded fuel rods from undiluted thermosetting binders filled with powdered graphite were unsuccessful, because of damage to coatings on fuel particles that resulted from strong particle-to-matrix bonding in conjunction with large matrix shrinkage on carbonization and subsequent irradiation. These problems have now been overcome through the use of a diluted thermosetting matrix with a low-char-yield additive (fugitive), which produces a more porous char similar to that from the pitch-based thermoplastic used in fabrication of fresh fuel. A 1-to-1 dilution of resin with fugitive produced the optimum binder for injection and carbonization, where the fired matrix in such rods contained about 20 wt% binder char and 80 wt% powdered graphite. Thermosetting fuel rods diluted with various amounts of fugitive to give binder chars that range from 12 to 48 wt% of the fired matrix have been subjected to irradiation screening tests, and rods with no more than 32 wt% binder char appear to perform about as well under irradiation as do pitch-based rods. However, particle damage does begin to occur in those lightly diluted rods in which the less-stable binder char constitutes more than 32 wt% of the fired matrix. (author)

Multi-scale modeling of the thermo-mechanical behavior of particle-based composites

International Nuclear Information System (INIS)

Di Paola, F.

2010-01-01

The aim of this work was to perform numerical simulations of the thermal and mechanical behavior of a particle-based nuclear fuel. This is a refractory composite material made of UO 2 spherical particles which are coated with two layers of pyrocarbon and embedded in a graphite matrix at a high volume fraction (45%). The objective was to develop a multi-scale modeling of this composite material which can estimate its mean behavior as well as the heterogeneity of the local mechanical variables. The first part of this work was dedicated to the modeling of the microstructure in 3D. To do this, we developed tools to generate random distributions of spheres, meshes and to characterize the morphology of the microstructure towards the finite element code Cast3M. A hundred of numerical samples of the composite were created. The second part was devoted to the characterization of the thermo-elastic behavior by the finite element modeling of the samples. We studied the influence of different modeling parameters, one of them is the boundary conditions. We proposed a method to vanish the boundary conditions effects from the computed solution by analyzing it on an internal sub-volume of the sample obtained by erosion. Then, we determined the effective properties (elastic moduli, thermal conductivity and thermal expansion) and the stress distribution within the matrix. Finally, in the third part we proposed a multi-scale modeling to determine the mean values and the variance and covariance of the local mechanical variables for any macroscopic load. This statistical approach have been used to estimate the intra-phase distribution of these variables in the composite material. (author) [fr

Multi-scale modeling of the thermo-mechanical behavior of particle-based composites

International Nuclear Information System (INIS)

Di Paola, F.

2010-11-01

The aim of this work was to perform numerical simulations of the thermal and mechanical behavior of a particle-based nuclear fuel. This is a refractory composite material made of UO 2 spherical particles which are coated with two layers of pyrocarbon and embedded in a graphite matrix at a high volume fraction (45 %). The objective was to develop a multi-scale modeling of this composite material which can estimate its mean behavior as well as the heterogeneity of the local mechanical variables. The first part of this work was dedicated to the modeling of the microstructure in 3D. To do this, we developed tools to generate random distributions of spheres, meshes and to characterize the morphology of the microstructure towards the finite element code Cast3M. A hundred of numerical samples of the composite were created. The second part was devoted to the characterization of the thermo-elastic behavior by the finite element modeling of the samples. We studied the influence of different modeling parameters, one of them is the boundary conditions. We proposed a method to vanish the boundary conditions effects from the computed solution by analyzing it on an internal sub-volume of the sample obtained by erosion. Then, we determined the effective properties (elastic moduli, thermal conductivity and thermal expansion) and the stress distribution within the matrix. Finally, in the third part we proposed a multi-scale modeling to determine the mean values and the variance and covariance of the local mechanical variables for any macroscopic load. This statistical approach have been used to estimate the intra-phase distribution of these variables in the composite material. (author)

Isolation and characterization of a homogeneous isoenzyme of wheat germ acid phosphatase.

Science.gov (United States)

Waymack, P P; Van Etten, R L

1991-08-01

An acid phosphatase (orthophosphoric monoester phosphohydrolase, acid optimum; EC 3.1.3.2) isoenzyme from wheat germ was purified 7000-fold to homogeneity. The effect of wheat germ sources and their relationship to the isoenzyme content and purification behavior of acid phosphatases was investigated. Extensive information about the purification and stabilization of the enzyme is provided. The instability of isoenzymes in the latter stages of purification appeared to be the result of surface inactivation together with a sensitivity to dilution that could be partially offset by addition of Triton X-100 during chromatographic procedures. Added sulfhydryl protecting reagents had no effect on activity or stability, which was greatest in the pH range 4-7. The purified isoenzyme was homogeneous by polyacrylamide gel electrophoresis and exhibited the highest specific activity and turnover number reported for any acid phosphatase. The molecular weights of the pure isoenzyme and of related isoenzymes from wheat germ were found to be identical (58,000). The pure isoenzyme contained a single polypeptide chain and had a negligible carbohydrate content. The amino acid composition was determined. Of the various reasons that were considered to explain isoenzyme occurrence, a genetic basis was considered most likely. The enzyme was found to exhibit substrate inhibition with some substrates below pH 6, while above pH 8 it exhibited downwardly curving Lineweaver-Burk plots of the type that are generally described as "substrate activation". The observation of a phosphotransferase activity was consistent with the formation of a covalent phosphoenzyme intermediate, while inactivation by diethyl pyrocarbonate was consistent with the presence of an active site histidine.

Coated particle fuel for high temperature gas cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Verfondern, Karl; Nabielek, Heinz [Research Center Julich (FZJ), Julich (Germany); Kendall, James M. [Global Virtual L1c, Prescott (United States)

2007-10-15

applications at 850-900 .deg. C and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 {mu}m diameter UO{sub 2} kernel of 10% enrichment is surrounded by a 100 {mu}m thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 {mu}m thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level.

Coated particle fuel for high temperature gas cooled reactors

International Nuclear Information System (INIS)

Verfondern, Karl; Nabielek, Heinz; Kendall, James M.

2007-01-01

and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO 2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

Detection and Analysis of Particles with Failed SiC in AGR-1 Fuel Compacts

International Nuclear Information System (INIS)

Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M.; Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A.

2014-01-01

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of "1"3"4Cs and "1"3"7Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during the AGR-1 irradiation test or post-irradiation safety testing at 1600– 1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the ORNL Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking. (author)

The significance of strength of silicon carbide for the mechanical integrity of coated fuel particles for HTRs

International Nuclear Information System (INIS)

Bongartz, K.; Scheer, A.; Schuster, H.; Taeuber, K.

1975-01-01

Silicon carbide (SiC) and pyrocarbon are used as coating material for the HTR fuel particles. The PyC shell having a certain strength acts as a pressure vessel for the fission gases whereas the SiC shell has to retain the solid fission products in the fuel kernel. For measuring the strength of coating material the so-called Brittle Ring Test was developed. Strength and Young's modulus can be measured simultaneously with this method on SiC or PyC rings prepared out of the coating material of real fuel particles. The strength measured on the ring under a certain stress distribution which is characteristic for this method is transformed with the aid of the Weibull formalism for brittle fracture into the equivalent strength of the spherical coating shell on the fuel particle under uniform stress caused by the fission gas pressure. The values measured for the strength of the SiC were high (400-700MN/m 2 ), it could therefore be assumed that a SiC layer might contribute significantly also to the mechanical strength of the fuel coating. This assumption was confirmed by an irradiation test on coated particles with PyC-SiC-PyC coatings. There were several particles with all PyC layers broken during the irradiation, whereas the SiC layers remained intact having to withstand the fission gas pressure alone. This fact can only be explained assuming that the strength of the SiC is within the range of the values measured with the brittle ring test. The result indicates that, in optimising the coating of a fuel particle, the PyC layers of a multilayer coating should be considered alone as prospective layers for the SiC. The SiC shell, besides acting as a fission product barrier, is then also responsible for the mechanical integrity of the particle

Graphitization kinetics of fluidized-bed pyrolytic carbons

International Nuclear Information System (INIS)

Beatty, R.L.

1975-08-01

Graphitization of 12 fluidized-bed pyrocarbons was studied as a function of heat-treatment time and temperature (1350 to 3000 0 C) to investigate the effect of initial microstructure on the graphitization process. The term ''graphitization'' is defined to include any thermally induced structural change, whether or not any layer stacking order is attained. A broad range of CVD microstructures was prepared at temperatures from 1150 to 1900 0 C and various propylene and methane concentrations. The twelve carbons spanned a wide range of graphitizabilities, primarily as a function of deposition temperature. Hydrocarbon concentration was of much less importance except for deposition at 1900 0 C. Hydrogen content of the as-deposited carbons decreased with increasing temperature of deposition, and initial graphitization behavior of the low-temperature carbons appeared to be related to hydrogen content and evolution. Rates of change in the parameters varied widely throughout the range of heat-treatment times (HTt) and temperatures (HTT) for the different carbons showing differences between the more graphitizable or ''soft'' carbons from the nongraphitizing or ''hard'' carbons. ΔH for nongraphitizing carbons was 175 +- 15 kcal below 1950 0 C, 240 +- 35 kcal at 1950 to 2700 0 C, and 330 +- 20 kcal above 2700 0 C. For graphitizing carbons deposited at 1150 0 C, values near 245 kcal were obtained from anti chi data for the HTT range 1350 to 1650 0 C, while densification data yielded values of about 160 kcal in the same range. The behaviors observed for graphitizable carbons above 2000 0 C are consistent with literature. Different kinetic behaviors below 2000 0 C were shown to be due to different initial microstructures as well as to different parameters measured. (U.S.)

Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

International Nuclear Information System (INIS)

Benchrif, A.; Chetaine, A.; Amsil, H.

2013-01-01

Highlights: • AFPR-100 reactor considered as a small nuclear reactor without on-site refueling originally based on TRISO micro-fuel element. • The AFPR-100 reactor was re-designed using the new Spherical Cermet fuel element. • The adoption of the Cermet fuel instead of TRISO fuel reduces the core lifetime operation by 3.1 equivalent full power years. • We discussed the new micro-fuel element candidate for small and medium sized reactors. - Abstract: The Atoms For Peace Reactor (AFPR-100), as a 100 MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY

Electrochemical performance of LiFePO4 modified by pressure-pulsed chemical vapor infiltration in lithium-ion batteries

International Nuclear Information System (INIS)

Li Jianling; Suzuki, Tomohiro; Naga, Kazuhisa; Ohzawa, Yoshimi; Nakajima, Tsuyoshi

2007-01-01

Using the pressure-pulsed chemical vapor infiltration (PCVI) technique, pyrolytic carbon (pyrocarbon) films were deposited on the surface of LiFePO 4 particles for cathode material of lithium-ion batteries. The electrochemical performance of the original LiFePO 4 and PCVIed LiFePO 4 materials was evaluated using a three electrodes cell by galvanostatic charging/discharging at 25, 40 and 55 deg. C, respectively. Morphology and structure of LiFePO 4 were analyzed by SEM, XRD and Raman. The resulting carbon contents at 500, 1000, 2000, 3000 and 5000 pulses were 2.7, 4.7, 9.5, 15.1 and 19.4%, respectively and these samples were abbreviated as 500P, 1000P, 2000P, 3000P and 5000P, respectively. All the PCVIed samples exhibited excellent rate performance. The tendency was more and more obvious with the increase of the current densities. The specific capacities of 500P, 1000P and 2000P were maintained at 117, 124 and 132 mAh g -1 , respectively, which were 120.8, 264.7 and 29.47% larger than those of corresponding original LiFePO 4 , respectively, at a 5C rate at 55 deg. C. The EIS measurement showed that electrochemical reaction resistance (R ct ) of PCVIed LiFePO 4 were obviously decreased, indicating a fast kinetics compared to the original LiFePO 4 . The cycle ability of the 2000P sample was tested at 25 deg. C and C/2 rate. The cell was cycled for 150 cycles and no obviously capacity fade was observed. Its specific capacity of 115 mAh g -1 at 150th cycle is 1.7 times higher than that of original LiFePO 4

Effect of Interface Modified by Graphene on the Mechanical and Frictional Properties of Carbon/Graphene/Carbon Composites

Science.gov (United States)

Yang, Wei; Luo, Ruiying; Hou, Zhenhua

2016-01-01

In this work, we developed an interface modified by graphene to simultaneously improve the mechanical and frictional properties of carbon/graphene/carbon (C/G/C) composite. Results indicated that the C/G/C composite exhibits remarkably improved interfacial bonding mode, static and dynamic mechanical performance, thermal conductivity, and frictional properties in comparison with those of the C/C composite. The weight contents of carbon fibers, graphene and pyrolytic carbon are 31.6, 0.3 and 68.1 wt %, respectively. The matrix of the C/G/C composite was mainly composed of rough laminar (RL) pyrocarbon. The average hardness by nanoindentation of the C/G/C and C/C composite matrices were 0.473 and 0.751 GPa, respectively. The flexural strength (three point bending), interlaminar shear strength (ILSS), interfacial debonding strength (IDS), internal friction and storage modulus of the C/C composite were 106, 10.3, 7.6, 0.038 and 12.7 GPa, respectively. Those properties of the C/G/C composite increased by 76.4%, 44.6%, 168.4% and 22.8%, respectively, and their internal friction decreased by 42.1% in comparison with those of the C/C composite. Owing to the lower hardness of the matrix, improved fiber/matrix interface bonding strength, and self-lubricating properties of graphene, a complete friction film was easily formed on the friction surface of the modified composite. Compared with the C/C composite, the C/G/C composite exhibited stable friction coefficients and lower wear losses at simulating air-plane normal landing (NL) and rejected take-off (RTO). The method appears to be a competitive approach to improve the mechanical and frictional properties of C/C composites simultaneously. PMID:28773613

Progress on Fabrication of Planar Diffusion Couples with Representative TRISO PyC/SiC Microstructure

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jolly, Brian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Campbell, Anne A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-10-01

Release of fission products from tristructural-isotropic (TRISO) coated particle fuel limits the fuel’s operational lifetime and creates potential safety and maintenance concerns. A need for diffusion analysis in representative TRISO layers exists to provide fuel performance models with high fidelity data to improve fuel performance and efficiency. An effort has been initiated to better understand fission product transport in, and release from, quality TRISO fuel by investigating diffusion couples with representative pyrocarbon (PyC) and silicon carbide (SiC). Here planar PyC/SiC diffusion couples are being developed with representative PyC/SiC layers using a fluidized bed chemical vapor deposition (FBCVD) system identical to those used to produce laboratory-scale TRISO fuel for the Advanced Gas Reactor Fuel Qualification and Development Program’s (AGR) first fuel irradiation. The diffusivity of silver, the silver and palladium system, europium, and strontium in the PyC/SiC will be studied at elevated temperatures and under high temperature neutron irradiation. The study also includes a comparative study of PyC/SiC diffusion couples with varying TRISO layer properties to understand the influence of SiC microstructure (grain size) and the PyC/SiC interface on fission product transport. The first step in accomplishing these goals is the development of the planar diffusion couples. The diffusion couple construction consists of multiple steps which includes fabrication of the primary PyC/SiC structures with targeted layer properties, introduction of fission product species and seal coating to create an isolated system. Coating development has shown planar PyC/SiC diffusion couples with similar properties to AGR TRISO fuel can be produced. A summary of the coating development process, characterization methods, and status are presented.

Uranium dispersion in the coating of weak-acid-resin-deprived HTGR fuel microspheres

International Nuclear Information System (INIS)

Weber, G.W.; Beatty, R.L.; Tennery, V.J.; Lackey, W.J. Jr.

1976-02-01

The current reference HTGR recycle fuel particle is a UO 2 /UC 2 kernel with a Triso coating comprising a low-density pyrocarbon (PyC) buffer, a high-density PyC inner LTI coating, SiC, and a high-density PyC outer LTI. The kernel is fabricated from a weak-acid ion exchange resin (WAR). Microradiographic examination of coated WAR particles has demonstrated that considerable U can be transferred from the kernel to the buffer coating during fabrication. Investigation of causes of fuel dispersion has indicated several different factors that contribute to fuel redistribution if not properly controlled. The presence of a nonequilibrium UC/sub 1-x/O/sub x/ (0 less than or equal to x less than or equal to 0.3) phase had no significant effect on initiating fuel dispersion. Gross exposure of the completed fuel kernel to ambient atmosphere or to water vapor at room temperature produced very minimal levels of dispersion. Exposure of the fuel to perchloroethylene during buffer and inner LTI deposition produced massive redistribution. Fuel redistribution observed in Triso-coated particles results from permeation of the inner LTI by HCl during SiC deposition. As the decomposition of CH 3 Cl 3 Si is used to deposit SiC, chlorine is readily available during this process. The permeability of the inner LTI coating has a marked effect on the extent of this mode of fuel dispersion. LTI permeability was determined by chlorine leaching studies to be a strong function of density, coating gas dilution, and coating temperature but relatively unaffected by application of a seal coat, variations in coating thickness, and annealing at 1800 0 C. Mechanical attrition of the kernels during processing was identified as a potential source of U-bearing fines that may be incorporated into the coating in some circumstances

Preparation and mass spectrometrical high temperature investigations on compounds of the quasi-ternary system Cs2O-Al2O3-SiO2

International Nuclear Information System (INIS)

Odoj, R.; Hilpert, K.; Nuernberg, H.W.

1977-09-01

Additions of aluminium oxide and silicen oxide to ceramic fuel for pyrocarbon-coated nuclear fuel paticles counteract a release of fission-cesium by compound formation. The vapourization tests carried out here using samples from the quasi-ternary system cesium-oxide-aluminium-oxide-silicon-oxide by means of high-temperature mass spectroscopy using a Knudsen cell served the optimization of this retention effect. The aim of the apparative changes on the knudsen cell were to shield heat radiation on the temperature measuring borehole through the tungsten wire cathode in order to be able to perform exact temperature measurements even below 1,000 0 C. A new method of preparation was developed to obtain defined cesium aluminium silicates whose composition was determined by Guinier and goniometer pictures as well as by microscopic investigations. According to the latter, 3 ternary compounds are present in the system investigated: CsAlSiO 4 , CsAlSi 2 O 6 and CsAlSi 5 O 12 . Their lattice constants were determined from goniometric measurements; the vapour pressure equection were set up from the measured cesium vapour pressure values over each sample and the enthalpies of the vapourization reactions were found to be 84 kcal for CsAlSiO 4 at 1,400 0 K, 100 kcal for CsAlSi 2 O 6 at 1,550 0 K and 122 kcal for CsAlSi 5 O 12 at 1,650 0 K. The cesium vapour pressures of the glas phases investigated of the system are above the Cs partial pressures of the solid crystalline phases of the same composition. The results of the work explain the causes of the reduction of the Cs release and show that the vapour pressure can be lowered by more than 10 orders of magnitude at reactor relevant temperatures by compound formation. (RB) [de

Electrochemical performance of LiFePO{sub 4} modified by pressure-pulsed chemical vapor infiltration in lithium-ion batteries

Energy Technology Data Exchange (ETDEWEB)

Li Jianling [Department of Physical Chemistry, University of Science and Technology Beijing, No. 30 College Road, Haidian District, Beijing 100083 (China); Department of Applied Chemistry, Aichi Institute of Technology, Yachigusa 1247, Yakusa-cho, Toyota 470-0392 (Japan)], E-mail: lijianling@metall.ustb.edu.cn; Suzuki, Tomohiro; Naga, Kazuhisa; Ohzawa, Yoshimi; Nakajima, Tsuyoshi [Department of Applied Chemistry, Aichi Institute of Technology, Yachigusa 1247, Yakusa-cho, Toyota 470-0392 (Japan)

2007-09-25

Using the pressure-pulsed chemical vapor infiltration (PCVI) technique, pyrolytic carbon (pyrocarbon) films were deposited on the surface of LiFePO{sub 4} particles for cathode material of lithium-ion batteries. The electrochemical performance of the original LiFePO{sub 4} and PCVIed LiFePO{sub 4} materials was evaluated using a three electrodes cell by galvanostatic charging/discharging at 25, 40 and 55 deg. C, respectively. Morphology and structure of LiFePO{sub 4} were analyzed by SEM, XRD and Raman. The resulting carbon contents at 500, 1000, 2000, 3000 and 5000 pulses were 2.7, 4.7, 9.5, 15.1 and 19.4%, respectively and these samples were abbreviated as 500P, 1000P, 2000P, 3000P and 5000P, respectively. All the PCVIed samples exhibited excellent rate performance. The tendency was more and more obvious with the increase of the current densities. The specific capacities of 500P, 1000P and 2000P were maintained at 117, 124 and 132 mAh g{sup -1}, respectively, which were 120.8, 264.7 and 29.47% larger than those of corresponding original LiFePO{sub 4}, respectively, at a 5C rate at 55 deg. C. The EIS measurement showed that electrochemical reaction resistance (R{sub ct}) of PCVIed LiFePO{sub 4} were obviously decreased, indicating a fast kinetics compared to the original LiFePO{sub 4}. The cycle ability of the 2000P sample was tested at 25 deg. C and C/2 rate. The cell was cycled for 150 cycles and no obviously capacity fade was observed. Its specific capacity of 115 mAh g{sup -1} at 150th cycle is 1.7 times higher than that of original LiFePO{sub 4}.

SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

Energy Technology Data Exchange (ETDEWEB)

Marshall, Douglas W.

2017-07-01

High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must have very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on the

High-quality thorium TRISO fuel performance in HTGRs

Energy Technology Data Exchange (ETDEWEB)

Verfondern, Karl [Forschungszentrum Juelich GmbH (Germany); Allelein, Hans-Josef [Forschungszentrum Juelich GmbH (Germany); Technische Hochschule Aachen (Germany); Nabielek, Heinz; Kania, Michael J.

2013-11-01

Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10{sup 25} n/m{sup 2} (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10{sup -5} for manufacturing, 2 x 10{sup -4} for normal operating conditions, and 5 x 10{sup -4

Properties Of A Midgut Trypanolysin From The Tsetse Fly Glossina Morsitans Morsitans

Directory of Open Access Journals (Sweden)

Mahamat H.Abakar

2015-08-01

Full Text Available The properties of a bloodmeal-induced trypanolysin from the midgut of the tsetse G. m. morsitans was studied in vitro. The semi-purified trypanolysin from twice-fed tsetse had the highest trypanolysin activity against bloodstream trypanosomes followed by those once-fed and the unfed flies. Serum found to display trypanolysin activity. The trypanolysin had no trypsin activity nor even affected by the enzyme. In addition trypanolysin was not affected by protease inhibitors such as soy bean trypsin inhibitor STI N-a-p-Tosyl-L-lysine chromethyl ketone TLCK phenylmethyl sulphonyl fluoride PMSF diisopropyl fluoro-phosphate DFP and tosylamide-2-phenylethyl chloromethyl ketone TPCK. However the activity was completely inhibited by diethyl pyrocarbonate DEPC and partially by aprotinin. The induction of trypanolysin activity by bloodmeal increased gradually reaching a peak at 72-120 h after the bloodmeal and then decreased rapidly with only 25 of the peak activity remaining after 192 h. The trypanolysin was inactivated during storage at 27amp8451 and 4amp8451 after 15 and 32 days respectively. Similarly heating the midguts trypanolysin to 60 - 80amp8451 led to loss of activity. On the other hand 50amp8451 was found to be the optimum temperature for trypanolysin activity. The activity was also unstable by freeze-thaw at 80amp8451 -70amp8451 -20amp8451 and 0amp8451 after 33 41 55 and 63 days respectively. Trypanolysin caused lyses of bloodstream-form T. b. brucei while the procyclic trypanosomes were unaffected. The highest trypanolysin activity in different tsetse species was found with Glossina longipennis followed by Glossina pallidipes Glossina morsitans centralis Glossina fuscipes fuscipes and G. m. morsitans. When the midgut homogenate was separated by anion-exchange chromatography the trypanolysin activity was recovered in the bound fraction. These results suggest that the midgut trypanolysin plays an important role in the establishment of

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D., E-mail: hunnjd@ornl.gov [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M. [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A. [Idaho National Laboratory (INL), P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-09-15

Highlights: • Cesium release was used to detect SiC failure in HTGR fuel. • Tristructural-isotropic particles with SiC failure were isolated by gamma screening. • SiC failure was studied by X-ray tomography and SEM. • SiC degradation was observed after irradiation and subsequent safety testing. - Abstract: As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of {sup 134}Cs and {sup 137}Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were

Thorium fuel performance assessment in HTRs

Energy Technology Data Exchange (ETDEWEB)

Allelein, H.-J. [Forschungszentrum Jülich, D-52425 Jülich (Germany); RWTH Aachen, D-52072 Aachen (Germany); Kania, M.J.; Nabielek, H. [Forschungszentrum Jülich, D-52425 Jülich (Germany); Verfondern, K., E-mail: k.verfondern@fz-juelich.de [Forschungszentrum Jülich, D-52425 Jülich (Germany)

2014-05-01

Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10{sup 25} m{sup −2} (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10{sup −5} for manufacturing; <2 × 10{sup −4} for normal operating conditions; and <5 × 10{sup −4} for accident conditions. These

High-quality thorium TRISO fuel performance in HTGRs

International Nuclear Information System (INIS)

Verfondern, Karl; Allelein, Hans-Josef; Nabielek, Heinz; Kania, Michael J.

2013-01-01

Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O 2 TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10 25 n/m 2 (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10 -5 for manufacturing, 2 x 10 -4 for normal operating conditions, and 5 x 10 -4 for accident conditions. These

Purification and characterization of an N alpha-acetyltransferase from Saccharomyces cerevisiae.

Science.gov (United States)

Lee, F J; Lin, L W; Smith, J A

1988-10-15

the most potent inhibitors. The enzyme was inactivated by chemical modification with diethyl pyrocarbonate and N-bromosuccinimide.

The known two types of transglutaminases regulate immune and stress responses in white shrimp, Litopenaeus vannamei.

Science.gov (United States)

Chang, Chin-Chyuan; Chang, Hao-Che; Liu, Kuan-Fu; Cheng, Winton

2016-06-01

Transglutaminases (TGs) play critical roles in blood coagulation, immune responses, and other biochemical functions, which undergo post-translational remodeling such as acetylation, phosphorylation and fatty acylation. Two types of TG have been identified in white shrimp, Litopenaeus vannamei, and further investigation on their potential function was conducted by gene silencing in the present study. Total haemocyte count (THC), differential haemocyte count (DHC), phenoloxidase activity, respiratory bursts (release of superoxide anion), superoxide dismutase activity, transglutaminase (TG) activity, haemolymph clotting time, and phagocytic activity and clearance efficiency to the pathogen Vibrio alginolyticus were measured when shrimps were individually injected with diethyl pyrocarbonate-water (DEPC-H2O) or TG dsRNAs. In addition, haemolymph glucose and lactate, and haemocytes crustin, lysozyme, crustacean hyperglycemic hormone (CHH), transglutaminaseI (TGI), transglutaminaseII (TGII) and clotting protein (CP) mRNA expression were determined in the dsRNA injected shrimp under hypothermal stress. Results showed that TG activity, phagocytic activity and clearance efficiency were significantly decreased, but THC, hyaline cells (HCs) and haemolymph clotting time were significantly increased in the shrimp which received LvTGI dsRNA and LvTGI + LvTGII dsRNA after 3 days. However, respiratory burst per haemocyte was significantly decreased in only LvTGI + LvTGII silenced shrimp. In hypothermal stress studies, elevation of haemolymph glucose and lactate was observed in all treated groups, and were advanced in LvTGI and LvTGI + LvTGII silenced shrimp following exposure to 22 °C. LvCHH mRNA expression was significantly up-regulated, but crustin and lysozyme mRNA expressions were significantly down-regulated in LvTGI and LvTGI + LvTGII silenced shrimp; moreover, LvTGII was significantly increased, but LvTGI was significantly decreased in LvTGI silenced shrimp

Irradiation performance of AGR-1 high temperature reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

2016-09-15

Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

International Nuclear Information System (INIS)

Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

2009-01-01

generally in the fuel element (pebble or compact) aims at estimating the source term of the fission products release outside the fuel element in normal operation or accidental conditions. In ATLAS, the transport mechanisms are modelled in a single transport law using effective diffusion coefficients for the fission product species in the different constitutive materials. The Verification and Validation of ATLAS code rests on two main steps: - Testing plan on fuel particle thermal mechanical behaviour has been carried out regarding sensitivity on dimensional parameters and physical properties such as kernel diameter, density and layer thicknesses and pyrocarbon layer anisotropy. The obtained results allow justifying and specifying the design for the manufacture. - Regarding fission product release under core heat-up accident conditions, the IAEA Coordinated Research Project 6 on 'Advanced in HTGR Fuel Technology Development' benchmark is the basis of the ATLAS code verification step. The ATLAS results obtained on IAEA benchmark cases with analytical solutions demonstrate that the models used fit the physical, chemical and mathematical laws. Regarding past irradiation tests and heating tests, ATLAS results show good agreement with the experimental database measurements. Comparison between ATLAS code results with analytical and experimental data allows defining confidence zones where ATLAS code gives accurate results and critical limits. These limits show where R and D efforts on models and material properties are needed to refine laws and models. (author)

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

Energy Technology Data Exchange (ETDEWEB)

Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

Energy Technology Data Exchange (ETDEWEB)

Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

International Nuclear Information System (INIS)

Boer, B.; Sen, R.S.; Pope, M.A.; Ougouag, A.M.

2011-01-01

The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO 2 kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 (micro)mm kernel diameter, 100 (micro)mm buffer, 35 (micro)mm IPyC, 35 (micro)mm SiC, 40 (micro)mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10 -2 failure probability. For a 'best-estimate' FGR fraction of 50

Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

International Nuclear Information System (INIS)

Marmier, Alain

2012-01-01

This thesis covers 3 fundamental aspects of High Temperature Reactor (HTR) performance: fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable future high-tech process heat applications with minimized R and D. The HTR concept features important inherent and passive safety characteristics: high thermal inertia and good thermal conductivity of the core; a negative Doppler coefficient; high quality of fuel elements and low power density. These features keep the core temperature within safe boundaries and minimise fission product release, even in case of severe accidents. The Very High Temperature reactor (VHTR) is based on the same safety concept as the initial HTR, but it aims at offering better economy with a higher reactor outlet temperature (and thus efficiency) and a high fuel discharge burn-up (and thus better sustainability). The inherent safety features of HTR have been demonstrated in small pebble-bed reactors in practice, but have to be replicated for reactors with industrially relevant size and power. An increase of the power density (in order to increase the helium coolant outlet temperature) leads to higher fuel temperatures and therefore higher fuel failure probability. The core of a pebble-bed reactor consists of 6 cm diameter spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. These pebbles contain thousands of 1 mm diameter fuel particles baked into a graphite matrix. These fuel particles, in turn, consist of a fuel kernel with successive coatings of pyrocarbon and silicon carbide layers. The coating layers are designed to contain the fission products that build up during operation of the reactor. The feasibility and performance of the fuel requires experimental verification in view of fuel qualification and licensing. For HTR fuel, the required test string comprises amongst others

Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

International Nuclear Information System (INIS)

Boer, Brian

2007-01-01

The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

The Role of Non-Destructive Testing in the Los Alamos Reactor Programme; Role des Essais Non Destructifs dans le Programme de Reacteurs de los Alamos; Rol' nedestruktivnykh ispytanij materialov v Los-Alamosskoj reaktornoj programme; Papel de los Metodos de Ensayo No Destructivo en el Programa de Reactores de Los Alamos

Energy Technology Data Exchange (ETDEWEB)

Tenney, G. H. [University of California, Los Alamos Scientific Laboratory, Los Alamos, NM (United States)

1965-10-15

The Los Alamos scientific Laboratory, operated Dy me University of california for me united states Atomic Energy Commission, has been active for more than twenty years in developing, designing, and building nuclear reactors of four general types: research, power, rocket propulsion, and critical assembly. The Non-destructive Testing Group serves practically all the activities and projects of the Laboratory; this paper describes some of the unique non-destructive testing techniques and applications developed for and used in the reactor programme. LAPRE (Los Alamos Power Reactor Experiment) was based on the use of a uranium phosphate solution at high temperature. This solution is very corrosive, and all parts in contact with it were clad with gold. Special radiographic techniques were used to inspect the gold during the process of producing rolled sheet from cast ingot. The welded seams were similarly inspected. An electrode-potential inspection method was developed for checking the gold surfaces for imbedded impurities. The fundamental concept of LAMPRE (Los Alamos Molten Plutonium Reactor Experiment) is the use of liquid - rather than solid - plutonium metal as fuel. Tantalum capsules contained the fuel. Novel nondestructive testing methods were used to check the soundness of base metal and welds during the production of the capsules, and to study the plutonium-loaded capsules before, during, and after melt-freeze tests. A molten plutonium pump experiment was followed with radiographic techniques, including a gamma-ray closed television circuit. For UHTREX (Ultra High Temperature Reactor Experiment), now under construction, micro radiographic and electron microscopic studies have been made on 150-{mu}m-diam. pyrocarbon-coated uranium carbide beads, to evaluate uranium migration as a function of temperature. The amount, and uniformity, of the uranium loading in the UHTREX graphite elements are determined with specially designed scintillation counters. About 90% of

13th International Conference on Films and Coatings

International Nuclear Information System (INIS)

2017-01-01

. For the first time this process was used to clean the surface of metals from radionuclides. The coefficient of purification at the level of 20 000 is the maximum compared to all existing methods. One of the new trends in the development of science and technology, reflected in the reports is the formation of pyrocarbon coatings in plasma of a vacuum arc discharge. For the first time such coating has been researched and applied to the electrodes of powerful generator tubes as antiemission coatings. Theoretically and experimentally were investigated the thermal processes during treatment of the inner surface of the cylindrical cavity by the cathode spots of a vacuum arc discharge. In a number of reports were reflected the characteristics of magnetron sputtering systems and principles of coatings deposition on their basis. The characteristics, technological aspects of production and results of testing of gradient coatings for aerospace optics were discussed. Promising technology of pulsed magnetron sputtering was noted. Possibilities of application of multilayer composite coatings in the systems of radiation protection of spacecraft were reviewed. Were shown the advantages of composite coatings before traditionally used in space technology aluminum alloys. At this Conference many reports were devoted to the formation of oxide coatings by different methods and for different fields of application. For example, the results of comparative studies of the original and processed in the plasma flow oxide microcomposites, consisting of TiO 2 , SiO 2 , Al 2 O 3 , and also plasma coatings from them – materials with amorphous-crystalline structure and a reinforced ultrafine phases of stishovite. It was shown that a reliable method of forming a specified surface nanorelief is a direct resistless lithography by a focused ion beam. The use of ions of different masses and energies significantly expands its abilities for nanoconstruction and nanoengineering of thin-layer structures