Nuclear reactor fuel element with a cluster of parallel fuel pins

International Nuclear Information System (INIS)

Macfall, D.; Butterfield, C.E.; Butterfield, R.S.

1977-01-01

An improvement of the design of nuclear reactor fuel elements is described and illustrated by the example of a gas-cooled, graphite-moderated nuclear reactor. The fuel element has a cluster of parallel fuel pins with an outer can of structure material and an inner sleeve, as well as tie bars and spacing devices for all of these parts. The fuel element designed according to the invention allows lasy assembling and disassembling before and after use. During use, no relative axial motions are possible; nevertheless, the graphite sleeve is at no time subject to tensile stress: the individual parts are held in position from below by a single holding device. (UWI) [de

Fabrication technology of spherical fuel element for HTR-10

International Nuclear Information System (INIS)

He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

2002-01-01

R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5

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

Natural uranium metallic fuel elements: fabrication and operating experience

International Nuclear Information System (INIS)

Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

1980-01-01

The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

Fuel element for high-temperature nuclear power reactors

International Nuclear Information System (INIS)

Schloesser, J.

1974-01-01

The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)

Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

Energy Technology Data Exchange (ETDEWEB)

Luo, Xiaowei, E-mail: xwluo@tsinghua.edu.cn; Wang, Xiaoxin; Shi, Li; Yu, Xiaoyu; Yu, Suyuan

2017-04-15

Highlights: • Graphite dust. • The wear properties of graphite. • Pebble bed. • High Temperature Gas-cooled Reactor. • Fuel element. - Abstract: The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10 MW High Temperature gas-cooled test Reactor (HTR-10).

Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

International Nuclear Information System (INIS)

Yang Lin; Liu Bing; Tang Chunhe

2014-01-01

Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

Science.gov (United States)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

Energy Technology Data Exchange (ETDEWEB)

Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

Distribution of fission products in Peach Bottom HTGR fuel element E01-01

International Nuclear Information System (INIS)

Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Fairchild, L.L.

1978-10-01

The fifth in a projected series of six postirradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements is described. The element analyzed received an equivalent of 897 full-power days of irradiation prior to the scheduled termination of Core 2 operation. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a 137 Cs inventory of 20.3 Ci in the graphite sleeve and 8.1 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides 134 Cs, /sup 110 m/Ag, 60 Co, and 154 Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the beta-emitters 3 H, 14 C, and 90 Sr were obtained at four axial locations of the fueled region of the element sleeve and two axial locations of the element spine. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. In addition to fission product distributions, the appearance of the component parts of the element was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed

Experimental study of some mounting brackets to support fuel elements

International Nuclear Information System (INIS)

Aubert, M.; Poglia, S.; Roche, R.

1958-09-01

In an atomic pile with vertical channels, fuel elements are stacked on one another. According to a possible assembly, fuel element can be contained by a graphite sleeve and be supported by a mounting bracket in this sleeve. Sleeves are then stacked on one another. The authors report the investigation of different designs for these mounting brackets. They describe their mechanical role and their mechanical, aerodynamic, neutronic and test conditions. They report tests performed on brackets made in graphite and on brackets made in stainless steel and graphite, and discuss the obtained results

Block fuel element for gas-cooled high temperature reactors

International Nuclear Information System (INIS)

Hrovat, M.F.

1978-01-01

The invention concerns a block fuel element consisting of only one carbon matrix which is almost isotropic of high crystallinity into which the coated particles are incorporated by a pressing process. This block element is produced under isostatic pressure from graphite matrix powder and coated particles in a rubber die and is subsequently subjected to heat treatment. The main component of the graphite matrix powder consists of natural graphite powder to which artificial graphite powder and a small amount of a phenol resin binding agent are added

Study on the Efficient Disintegration of HTGR Fuel Elements by Electrochemical Method

International Nuclear Information System (INIS)

Piao Nan; Chen Ji; Xiao Cuiping; We Mingfen; Che Jing

2014-01-01

The spent fuel elements in High- temperature gas-cooled reactor (HTGR) have a special structure, so the head-end process of the spent fuel reprocessing is different from the process of water reactor spent fuel. The first step of head-end process of the HTGR spent fuel reprocessing process is disintegration of the graphite matrix and separation of the coated fuel particles. Electrochemical method with nitrate solution as an electrolyte for fuel element disintegration has been conducted by the Institute of Nuclear and New Energy Technology in Tsinghua University. This method allows a total disintegration of graphite matrix, while still preserving the integrity of TRISO particles. The influences of the pretreatment methods such as heating oxidation of graphite, hydrothermal and oxidants oxidation were investigated in the present work. The experimental results showed that there were no significant effects on increasing the disintegration rate when pretreatment methods were used ahead of electrochemical disintegration. This phenomenon indicated that the fuel elements which were calcined at 1073 K and pressed under 300 MPa are too compact to be broken by these pretreatment methods. And the electrochemical disintegration is an effective but slow method in breaking the graphite matrix. (author)

FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

International Nuclear Information System (INIS)

Pierce, V.H.

2005-01-01

1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

Method of manufacturing a graphite coated fuel can

International Nuclear Information System (INIS)

Saito, Koichi; Uchida, Shunsuke.

1984-01-01

Purpose: To improve the close bondability and homogeneity of a graphite coating formed at the inner surface of a fuel can. Method: A coating containing graphite dispersed in a volatile organic solvent is used and a graphite coating is formed to the inner surface of a fuel can by way of a plunger method. After applying graphite coating, an inert gas is caused to flow at a certain flow rate to the inside of the fuel can horizontally rotaged so that gassification and evaporation of the volatile organic solvent contained in the graphite coating may be promoted. Since drying of the graphite coating coated to the inner surface of the fuel can thus be controlled, a graphite coating with satisfactory close bondability and homogeneity can be formed. (Kawakami, Y.)

Graphite for high-temperature reactors

International Nuclear Information System (INIS)

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

1976-01-01

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

Method of making a graphite fuel element having carbonaceous fuel bodies

International Nuclear Information System (INIS)

Miertschin, G.N.; Leary, D.F.

1977-01-01

Particulate nuclear fuel material, particulate carbon and pitch are combined with an additive which is effective to reduce the coke yield upon carbonization to mold a green fuel body. The additive may be polystyrene, a styrene-butadiene copolymer, an aromatic hydrocarbon having a molecular weight between about 75 and 300 or a saturated hydrocarbon polymer. The green fuel body is inserted in a complementary cavity within a porous nuclear fuel element body and heated in situ to decompose the pitch and additive, leaving a relatively close-fitting fuel body in the cavity

Fuel element structure - design, production and operational behaviour

International Nuclear Information System (INIS)

Pott, G.; Dietz, W.

1985-01-01

The lectures held at the meeting of the fuel element section of the Kerntechnische Gesellschaft gives a survey of developments in fuel element structure design for PWR-type, BWR-type and fast breeder reactors. For better utilization of the fuel, concepts have been developed for re-usable, removable and thus repairable fuel elements. Furthermore, the manufacturing methods for fuel element structures were refined to achieve better quality and more efficient manufacturing methods. Statements on the dimensional behaviour and on the mechanical stability of fuel element structures in normal and accident operation could be made on the basis of post-irradiation inspections. Finally, the design, manufacture and irradiation behaviour of graphite reflectors in HTGR-type reactors are described. The 12 lectures have been recorded in the data base separately. (RF) [de

Characterization of graphite-matrix pulsed reactor fuels

International Nuclear Information System (INIS)

Karnes, C.H.; Marion, R.H.

1976-01-01

The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

Long-term testing of HTR fuel elements in the Federal Republic of Germany

International Nuclear Information System (INIS)

Nickel, H.

1986-12-01

The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.) [de

Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

International Nuclear Information System (INIS)

Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

2013-01-01

The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

Development of a dry-mechanical graphite separation process and elimination of the separated carbon for the reprocessing of spherical HTR fuel elements

International Nuclear Information System (INIS)

Kronschnabel, H.

1982-01-01

Due to the C-14 distribution the separation of the particle-free outer region of the spherical HTR fuel element with subsequent solidification of the separated carbon makes it possible to reduce by half the remaining C-14 inventory in the inner particle region to be further treated. Separation of the particle-free outer region by a newly developed sphere-peeling milling machine, conditioning the graphite into compacts and in-situ cementation into a salt-mine are the basic elements of this head-end process variation. An annual cavern volume of approx. 2000 m 3 will be needed to ultimately store the graphite of the particle-free outer region, which corresponds to a reprocessing capacity of 50 GWsub(e) installed HTR power. The brush-disintegration of the remaining inner particle region and the resulting peel-brush-preparation are capable of separating 95% of the graphite without any heavy metal losses. With the mentioned reprocessing capacity an annual cavern volume of approx. 16.500 m 3 is required. (orig.) [de

Gas-cooled nuclear reactor with a filling of spherical fuel elements

International Nuclear Information System (INIS)

Hantke, H.J.

1978-01-01

In order to protect the reflector blanket of a pebble bed reactor against radiation damage a filling of graphite spheres is arranged between blanket and fuel elements, having got a smaller diameter than fuel spheres. Before reaching unduely high irradiation values caused by fast neutrons these graphite spheres are removed from the core, together with the usual discharge of spheres, and replaced by new spheres. (TK) [de

Preparation of spherical fuel elements for HTR-PM in INET

International Nuclear Information System (INIS)

Xiangwen, Zhou; Zhenming, Lu; Jie, Zhang; Bing, Liu; Yanwen, Zou; Chunhe, Tang; Yaping, Tang

2013-01-01

Highlights: • Modifications and optimizations in the manufacture of spherical fuel elements (SFE) for HTR-PM are presented. • A newly developed overcoater exhibits good stability and high efficiency in the preparation of overcoated particles. • The optimized carbonization process reduces the process time from 70 h in the period of HTR-10 to 20 h. • Properties of the prepared SFE and matrix graphite balls meet the design specifications for HTR-PM. • In particular the mean free uranium fraction of 5 consecutive batches is only 8.7 × 10 −6 . -- Abstract: The spherical fuel elements were successfully manufactured in the period of HTR-10. In order to satisfy the mass production of fuel elements for HTR-PM, several measures have been taken in modifying and optimizing the manufacture process of fuel elements. The newly developed overcoater system and its corresponding parameters exhibited good stability and high efficiency in the preparation of overcoated particles. The optimized carbonization process could reduce the carbonization time from more than 70 h to 20 h and improve the manufacturing efficiency. Properties of the manufactured spherical fuel elements and matrix graphite balls met the design specifications for HTR-PM. The mean free uranium fraction of 5 consecutive batches was 8.7 × 10 −6 . The optimized fuel elements manufacturing process could meet the requirements of design specifications of spherical fuel elements for HTR-PM

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.)

Influence of graphite discs, chamfers, and plenums on temperature distributions in high burnup fuel

International Nuclear Information System (INIS)

Ranger, A.; Tayal, M.; Singh, P.

1990-04-01

Previous studies have demonstrated the desirability to increase the fuel burnups in CANDU reactors from 7-9 GW.d/t to 21 GW.d/t. At high burnups, one consideration in fuel integrity is fission gas pressure, which is predicted to reach about 13 MPa. The gas pressure can be kept below the coolant pressure (about 10 MPa) via a variety of options such as bigger chamfers, deeper dishes, central hole, and plenums. However, it is important to address the temperature perturbations produced by the bigger chamfers and plenums which in turn, affect the gas pressure. Another consideration in fuel integrity is to reduce the likelihood of fuel failures via environmentally assisted cracking. Insertion of graphite discs between neighbouring pellets will lower the pellet temperatures, hence, lower fission gas release and lower expansion of the pellet. Therefore, it is desired to quantify the effect of graphite discs on pellet temperatures. Thermal analyses of different fuel element geometries: with and without chamfers, graphite discs, and plenums were performed. The results indicate that the two-dimensional distributions of temperatures due to the presence of chamfers, graphite discs, or plenums can have a significant impact on the integrity of high burnup fuel as we have been able to quantify in this paper

Postirradiation examination and evaluation of Peach Bottom fuel test element FTE-6

International Nuclear Information System (INIS)

Wallroth, C.F.; Holzgraf, J.F.; Jensen, D.D.

1977-09-01

Fuel test element FTE-6 was irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) for 645 equivalent full power days. Four fuel varieties, contained in H-327 graphite bodies, were tested. A primary result of this test has been to demonstrate acceptable performance even with calculated high stresses in the graphite bodies. Heterogeneous fuel loadings in the element caused local power peaking and azimuthal power variations, deforming the graphite fuel bodies and thereby causing bowing nearly five times as large as the diametral clearance within the sleeve. The axial stresses resulting from interference between the fuel bodies and sleeve were estimated to have reached 45% of the ultimate material strength at the end of the irradiation. Residual stresses from differential contraction within the fuel body resulted in probable in-plane stress levels of 130% of the material strength at the end-of-life shutdown and of up to 150% of the strength at shutdown during the irradiation cycle. The high in-plane stresses are local peaks at the corners of a sharp notch in the element, which may account for the stresses failing to cause damage. The lack of observable damage, however, indicates that the methods and data used for stress analysis give results that are either fairly accurate or conservative

The calculation - experimental investigations of the HTGR fuel element construction

International Nuclear Information System (INIS)

Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

1985-01-01

One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

Process for the production of fuel combined articles for addition in block shaped high temperature fuel elements

International Nuclear Information System (INIS)

Hrovat, M.; Rachor, L.

1976-01-01

There is provided a process for the production of fuel compacts consisting of an isotropic, radiation-resistant graphite matrix of good heat conductivity having embedded therein coated fuel and/or fertile particles for insertion into high temperature fuel elements by providing the coated fuel and/or fertile particles with an overcoat of molding mixture consisting of graphite powder and a thermoplastic resin binder. The particles after the overcoating are provided with hardener and lubricant only on the surface and subsequently are compressed in a die heated to a constant temperature of about 150 0 C, hardened and discharged therefrom as finished compacts

Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

International Nuclear Information System (INIS)

Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

2016-01-01

In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

Development and engineering plan for graphite spent fuels conditioning program

International Nuclear Information System (INIS)

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

1993-09-01

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

Fuel element reactivity worth in different rings of the IPR-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Gomes do Prado Souza, Rose Mary

2008-10-29

The thermal power of the IPR-R1 TRIGA Reactor will be upgraded from 100 kW to 250 kW. Starting core: loaded with 59 aluminum cladded fuel elements; 1.34 $ excess reactivity; and 100 kW power. It is planned to go 2.5 times the power licensed, i.e., 250 kW. This forces to enlarge the reactivity level. Nuclear reactors must have sufficient excess reactivity to compensate the negative reactivity feedback effects caused by: the fuel temperature, fuel burnup, fission poisoning production, and to allow full power operation for predetermined period of time. To provide information for the calculation of the new core arrangement, the reactivity worth of some fuel elements in the core were measured as well as the determination of the core reactivity increase in the substitution of the original fuels, cladded with aluminium, for new ones, cladded with stainless steel. The reactivity worth of fuel element was measured from the difference in critical position of the control rods, calibrated by the positive period method, before and after the fuel element was withdrawn from the core. The magnitude of reactivity increase was determined when withdrawing the original Al-clad fuel (a little burned up) and the graphite elements, and inserting a fresh Al-clad fuel element, one by one. Experimental results indicated that to obtain enough reactivity excess to increase the rector power the addition of 4 new fuel elements in the core would be sufficient: - Substitution of 4 Al-clad fuel elements in ring C for fresh stainless steel clad fuel elements; - increase the reactivity {approx_equal} 4 x 6.5 = 26 cents; - The removed 4 Al-clad F. E. (a little burned up) put in the core periphery, ring F, replacing graphite elements; - add < 4 x 39 156 cents (39 cents was measured with a fresh F.E.). Neutron source was changed from position F7 to F8. Control and Safety rods were moved from ring D to C in order to increase their reactivity worth. Regulating rod was kept at the same position, F16. Four

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

Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

International Nuclear Information System (INIS)

Verfondern, K.; Mueller, D.

1991-01-01

Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

Nuclear fuel element

International Nuclear Information System (INIS)

Knowles, A.N.

1979-01-01

A nuclear fuel-containing body for a high temperature gas cooled nuclear reactor is described which comprises a flat plate in which the nuclear fuel is contained as a dispersion of fission product-retaining coated fuel particles in a flat sheet of graphitic or carbonaceous matrix material. The flat sheet is clad with a relatively thin layer of unfuelled graphite bonded to the sheet by being formed initially from a number of separate preformed graphitic artefacts and then platen-pressed on to the exterior surfaces of the flat sheet, both the matrix material and the artefacts being in a green state, to enclose the sheet. A number of such flat plates are supported edge-on to the coolant flow in the bore of a tube made of neutron moderating material. Where a number of tiers of plates are superimposed on one another, the abutting edges are chamfered to reduce vibration. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Effect of various dopant elements on primary graphite growth

International Nuclear Information System (INIS)

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

2012-01-01

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

Gamma scanning of full scale HTR fuel elements

International Nuclear Information System (INIS)

Harrison, T.A.; Simpson, J.A.H.; Nabielek, H.

1983-04-01

Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

Nondestructive examination of 54 fuel and reflector elements from Fort St. Vrain core segment 2

International Nuclear Information System (INIS)

Saurwein, J.J.

1982-10-01

Fifty-four fuel and reflector elements irradiated in core segment 2 of the Fort St. Vrain high-temperature gas-cooled reactor (HTGR) were nondestructively examined. The time- and volume-averaged graphite irradiation temperatures for the elements ranged from approx. 350 0 to 750 0 C. The element-averaged fast neutron fluences ranged from approx. 0.2 to 1.6 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/. The elements, except for two fuel elements in which single localizeed cracks developed during irradiation, were in excellent condition. No evidence was observed of significant graphite oxidation or mechanical interaction beween elements. The cracks in the two elements did not affect their performance or handling. These elements were, otherwise, in excellent condition. Nearly all elements shrank in both the axial and radial directions, but the dimensional changes were relatively small

Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

International Nuclear Information System (INIS)

Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

1990-01-01

In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

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

International Nuclear Information System (INIS)

Gesh, Christopher J.

2004-01-01

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

Recent developments concerning French fuel elements used in natural uranium - graphite - CO{sub 2} reactor systems; Developpements recents des elements combustibles francais de la filiere uranium naturel - graphite - CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Salesse, M; Stohr, J A; Jeanpierre, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The policy followed in France for the development of fuel elements for reactors belonging to the Electricite de France has been to benefit as much as possible, for each new pile from the most recent technical progress by developing in each case a fuel element allowing the maximum power per channel. The two latest fuel elements thus studied by the French Atomic Energy Commission are of two different types: a tubular uranium element closed at both ends and cooled externally. (This type of element, chosen for the reactors EDF 2, EDF 3 and EDF 4 makes it possible to attain maximum specific powers of the order of 6 MW/metric ton.); an open tubular uranium element cooled both internally and externally, called an annular element which in being studied as a possibility for EDF 5. Such an element makes it possible to attain specific powers of over 12 MW/metric ton. The two types of element have the following common characteristics: - the can, for external cooling, has herning-bone type fins. This type of profile which has been vastly improved recently thereby increasing its thermal efficiency, has the important advantage of avoiding vibration of the element, but has posed problems of resistance to thermal cycling necessitating much research. - the fuel rods are placed inside graphite jackets, this limiting the vertical forces to which they are subjected and protecting them during charging and discharging. On the other hand, these elements present very different problems as for as the following points are concerned: - the characteristics required of the uranium tubes apart of course from a good dimensional stability during irradiation in the two cases are in the case of the closed tubes a very high resistance to external pressure, and in the case of the annular elements a low neutron absorption. Thus for each of these two cases it has been necessary to develop a suitable type of alloy. - a possible loosening of the can during thermal cycling, which is peculiar to the

Fission product release from HTGR coated microparticles and fuel elements

International Nuclear Information System (INIS)

Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

1991-01-01

The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

Determination of the hydrogen content of fuel elements

International Nuclear Information System (INIS)

Soare, M.; Petriu, F.; Toma, V.

1995-01-01

A new method and apparatus are reported for determination of the total hydrogen content by measurements on as-manufactured fuel elements, heated at prescribed temperature values between 200 degrees C and 600 degrees C. The method is based on the catalytic oxidation of the organic compounds and transformation of the hydrogen in the equivalent water quantity which is analysed by a special infrared detector. Different types of measurements for determination of the hydrogen content from graphite coating, UO 2 pellets and filling gas are presented. Also, experimental observation regarding water release and graphite thermal decomposition kinetic are discussed. (author)

Fort St. Vrain graphite site mechanical separation concept selection

International Nuclear Information System (INIS)

Berry, S.M.

1993-09-01

One of the alternatives to the disposal of the Fort St. Vrain (FSV) reactor spent nuclear fuel involves the separation of the fuel rods composed of compacts from the graphite fuel block assembly. After the separation of these two components, the empty graphite fuel blocks would be disposed of as a low level waste (provided the appropriate requirements are met) and the fuel compacts would be treated as high level waste material. This report deals with the mechanical separation aspects concerning physical disassembly of the FSV graphite fuel element into the empty graphite fuel blocks and fuel compacts. This report recommends that a drilling technique is the preferred choice for accessing the, fuel channel holes and that each hole is drilled separately. This report does not cover any techniques or methods to separate the triso fuel particles from the graphite matrix of the fuel compacts

The industrial production of fuel elements; La fabrication en france des elements combustibles

Energy Technology Data Exchange (ETDEWEB)

Boussard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Nadal, J [Societe Industrielle de Combustible Nucleaire (SICN), 75 - Paris (France); Pellen, A [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques (CERCA), 75 - Paris (France)

1964-07-01

The authors deal successively with the industrial production of fuel elements for power reactors of the natural uranium-graphite-gas type, and more particularly for the EDF power stations, and with the industrial production of fuel elements containing enriched uranium designed for swimming-pool type reactors. 1. part: advanced fuel elements for the EDF reactors. After recalling the characteristics of the fuel elements now being produced industrially for the Marcoule and Chinon reactors, the authors give the various steps leading to the industrial production of a new type of fuel element both as concerns the can, and in certain cases the graphite sleeve, and the fuel itself. As for as the production of the fuel is concerned, they describe the various operations, stressing the original aspects of the production and of the equipment such as: - casting in hot moulds, - thermal treatments, of Uranium containing 1% in weight molybdenum, - welding of the pellets for closing the tubes of uranium, - canning, - controls in the various steps. As far as can production is concerned they show why the extruded can was replaced by a machined can and give a few characteristics of the equipment used as well as the controls effected. They give also some details concerning the production and machining of the sleeves. After recalling the state of the nuclear fuel industry in France in mid-1964 the authors stress the economic aspects of the production of fuel elements. They show the relative importance of capital costs on the cost price of the fuel itself and examine the various items involved. They analyse the cost price of a completed fuel element using present date knowledge. In conclusion the authors show the particular points which should be the subject of future efforts in order to decrease the cost of a production which is perhaps delicate but now will define, and review the development of this new industrial branch. 2. part: industrial production of fuel elements for swimming

Improvements in the preparation of nuclear fuel elements with addition of a molding mixture to fuel particles

International Nuclear Information System (INIS)

Miertschin, G.N.; Leary, D.F.

1975-01-01

An improved molting mixture to be added to nuclear fuel particles for the preparation of nuclear fuel elements is presented. It consists of carbon and pitch particles and contains an additive reducing the final coke yield of the fuel mass formed. This additive is chosen from: polystyrene and copolymers of styrene and butadiene of molecular weight between 500 and 1000000; aromatic compounds of molecular weight between 75 and 300; saturated hydrocarbon polymers of molecular weight between 500 and 1000000. The additive may be camphor, naphthalene, anthracene, phenanthrene, dimethyl terephthalate or their mixtures and is present at a concentration of 5 to 50% by weight. The carbon particles used consist of powdered graphite. These fuel elements are intended for gas-cooled high-temperature reactors [fr

Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

International Nuclear Information System (INIS)

Weissfloch, R.

The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Weissfloch, R

1973-07-15

The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

Status and aspects of fuel element development for advanced high-temperature reactors in the FRG

International Nuclear Information System (INIS)

Nickel, H.; Balthesen, E.

1975-01-01

In the FRG three basic fuel element designs for application in high temperature gas cooled reactors are being persued: the spherical element, the graphite block element, and the moulded block element (monolith). This report gives the state of development reached with the three types of elements but also views their specific merits and performance margin and presents aspects of their future development potential for operation in advanced HTGR plants. The development of coated feed and breed particles for application in all HTGR fuel elements is treated in more detail. Summarizing it can be said that all the fuel elements as well as their components have proved their aptitude for the dual cycle systems in numerous fuel element and particle performance tests. To adapt these fuel elements and coated particles for advanced reactor concepts and to develop them up to full technical maturity further testing is still necessary, however. Ways of overcoming problems arising from the more stringent requirements are shown. (orig.) [de

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

Directory of Open Access Journals (Sweden)

J. Lacaze

2013-01-01

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

Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

Energy Technology Data Exchange (ETDEWEB)

Engle, G.B.; Johnson, W.R.

1976-10-08

Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

Neutronic calculation of the next fuel elements for the Argonaut reactor

International Nuclear Information System (INIS)

Oliveira, C.R.E.; Brito Aghina, L.O. de

1981-01-01

The best parameters of the next fuel elements of the Argonaut reactor, at IEN (Instituto de Engenharia Nuclear - Brazil), were determined. The next fuel elements will be rods of an uranium mixture (19.98% enriched), graphite and bakelite. The parameters to be determined are: mixture density, percentage of uranium in the mixture, pellet radius, rod material and elements arrangement (step). The calculations routines consisted in the analysis of several steps, using the LEOPARD computer code for cell calculations and RMAT1D for one dimensional spatial calculations (criticality) with four energy groups. Finally a neutronic study of the Argounat reactors present configuration was done, using the HAMMER computer code (cell), the EXTERMINATOR computer code (two-dimensional calculations) and RAMAT1D. (Author) [pt

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

Science.gov (United States)

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

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

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

International Nuclear Information System (INIS)

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

1988-06-01

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

Studies on design principles and criteria of fuels and graphites for experimental multi-purpose very high temperature reactor

International Nuclear Information System (INIS)

Arai, Taketoshi; Sato, Sadao; Tani, Yutaro

1977-12-01

Design principles and criteria of fuels and graphites have been studied to determine the main design parameters of a reference core MARK-III of the Experimental Multi-purpose Very High Temperature Reactor. The present status of research and development for HTGR fuels and graphites is reviewed from a standpoint of their integrity and safety aspects, and is compared to the specific design requirements for the VHTR fuels and graphites. Consequently, reasonable materials specifications, safety criteria and design analysis methods are presented for coated fuel particle, fuel compact, graphite sleeve, core support graphite and neutron absorber material. These design principles and criteria will be refined by further experimental investigations. (auth.)

Fuel management of HTR-10

International Nuclear Information System (INIS)

Wu Zongxin; Jing Xingqing

2001-01-01

The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

Improved graphite matrix for coated-particle fuel

International Nuclear Information System (INIS)

Schell, D.H.; Davidson, K.V.

1978-10-01

An experimental process was developed to incorporate coated fuel particles in an extruded graphite matrix. This structure, containing 41 vol% particles, had a high matrix density, >1.6 g/cm 3 , and a matrix conductivity three to four times that of a pitch-injected fuel rod at 1775 K. Experiments were conducted to determine the uniformity of particle loadings in extrusions. Irradiation specimens were supplied for five tests in the High-Fluence Isotope Reactor at the Oak Ridge National Laboratory

Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors; Etude de la chute thermique au contact uranium-gaine pour des elements combustibles de reacteur de la filiere graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Faussat, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Levenes, G; Michel, M [Societe Industrielle de Combustible Nucleaire (France)

1964-07-01

The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm{sup 2} seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors) [French] Le rapport fait le point des essais actuellement en cours au CEA pour determiner la resistance thermique de contact uranium-gaine pour des reacteurs de la filiere graphite-gaz. Ces essais sont effectues en laboratoire sur des appareils bases sur le principe d'une circulation de flux de chaleur a travers un empilement d'eprouvettes dont les faces en contact sont planes. De l'etude, il ressort essentiellement que: - pour un element a uranium metallique et gaine de magnesium type G-2 ou EdF-2, on peut admettre la valeur de 0,2 deg C/W/cm{sup 2} pour des temperatures de gaines de 400 deg C et plus. - cette valeur ne depend pas de l'etat de surface microgeometrique de l'uranium pour un domaine de rugosites couvrant largement celles que l'on observe sur des tubes et barreaux fabriques en serie. - pour les gaines internes d'elements a refroidissement interne et externe la valeur de la resistance de contact reste a preciser pour les temperatures inferieures a 400 deg C, en fonction des contraintes existant dans les

Development of synthetic graphite resistive elements for sintering furnace

International Nuclear Information System (INIS)

Otani, C.; Rezende, Mirabel C.; Polidoro, H.A.; Otani, S.

1987-01-01

The synthetic graphites have been produced using lignin coke, natural graphite and phenolic resin. The bulk density, porosity, flexural strength and eletrical resistivity measurements have been performed on specimens at about 2400 0 C. The performance of these materials, as heating elements, was evaluated in a sintering furnace prototype. This paper reports the fabrication process and the experimental results. (Author) [pt

Test of high temperature fuel element, (1)

International Nuclear Information System (INIS)

Akino, Norio; Shiina, Yasuaki; Nekoya, Shin-ichi; Takizuka, Takakazu; Emori, Koichi

1980-11-01

Heat transfer experiment to measure the characteristics of a VHTR fuel in the same condition of the reactor core was carried out using HTGL (High Temperature Helium Gas Loop) and its test section. In this report, the details of the test section, related problems of construction and some typical results are described. The newly developed heater with graphite heat transfer surface was used as a simulated fuel element to determine the heat transfer characteristics. Following conclusions were obtained; (1) Reynolds number between turbulent and transitional region is about 2600. (2) Reynolds number between transitional and laminar region is about 4800. (3) The laminarization phenomena have not been observed and are hardly occurred in annular tubes comparing with round tube. (4) Measured Nusselt numbers agree to the established correlations in turbulent and laminar regions. (author)

Corrosion of graphite composites in phosphoric acid fuel cells

Science.gov (United States)

Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

1986-01-01

Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

Development and testing of the EDF-2 reactor fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.

1964-01-01

This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

2013-01-01

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

2013-07-15

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

Nondestructive examination of 51 fuel and reflector elements from Fort St. Vrain Core Segment 1

International Nuclear Information System (INIS)

Miller, C.M.; Saurwein, J.J.

1980-12-01

Fifty-one fuel and reflector elements irradiated in core segment 1 of the Fort St. Vrain High-Temperature Gas-Cooled Reactor (HTGR) were inspected dimensionally and visually in the Hot Service Facility at Fort St. Vrain in July 1979. Time- and volume-averaged graphite temperatures for the examined fuel elements ranged from approx. 400 0 to 750 0 C. Fast neutron fluences varied from approx. 0.3 x 10 25 n/m 2 to 1.0 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/. Nearly all of the examined elements shrank in both axial and radial dimensions. The measured data were compared with strain and bow predictions obtained from SURVEY/STRESS, a computer code that employs viscoelastic beam theory to calculate stresses and deformations in HTGR fuel elements

Roughness analysis of graphite surfaces of casting elements

Directory of Open Access Journals (Sweden)

M. Wieczorowski

2010-01-01

Full Text Available In the paper profilometric measurements of graphite casting elements were described. Basic topics necessary to assess roughness of their surfaces and influence of asperities on various properties related to manufacturing and use were discussed. Stylus profilometer technique of surface irregularities measurements including its limits resulting from pickup geometry and its contact with measured object were ana-lyzed. Working principle of tactile profilometer and phenomena taking place during movement of a probe on a measured surface were shown. One of the important aspects is a flight phenomenon, which means movement of a pickup without contact with a surface during inspection resulting from too high scanning speed. results of comparison research for graphite elements of new and used mould and pin composing a set were presented. Using some surface roughness, waviness and primary profile parameters (arithmetical mean of roughness profile heights Ra, biggest roughness profile height Rz, maximum primary profile height Pt as well as maximum waviness profile height Wt a possibility of using surface asperities parameters as a measure of wear of chill graphite elements was proved. The most often applied parameter is Ra, but with a help of parameters from W and P family it was shown, that big changes occur not only for roughness but also for other components of surface irregularities.

Surface-reconstructed graphite nanofibers as a support for cathode catalysts of fuel cells.

Science.gov (United States)

Gan, Lin; Du, Hongda; Li, Baohua; Kang, Feiyu

2011-04-07

Graphite nanofibers (GNFs), on which surface graphite edges were reconstructed into nano-loops, were explored as a cathode catalyst support for fuel cells. The high degree of graphitization, as well as the surface-reconstructed nano-loops that possess topological defects for uniform metal deposition, resulted in an improved performance of the GNF-supported Pt catalyst.

Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

International Nuclear Information System (INIS)

Silva, Alice Cunha da; Su, Jian

2013-01-01

The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

Progress in Developing Finite Element Models Replicating Flexural Graphite Testing

International Nuclear Information System (INIS)

Bratton, Robert

2010-01-01

This report documents the status of flexural strength evaluations from current ASTM procedures and of developing finite element models predicting the probability of failure. This work is covered under QLD REC-00030. Flexural testing procedures of the American Society for Testing and Materials (ASTM) assume a linear elastic material that has the same moduli for tension and compression. Contrary to this assumption, graphite is known to have different moduli for tension and compression. A finite element model was developed and demonstrated that accounts for the difference in moduli tension and compression. Brittle materials such as graphite exhibit significant scatter in tensile strength, so probabilistic design approaches must be used when designing components fabricated from brittle materials. ASTM procedures predicting probability of failure in ceramics were compared to methods from the current version of the ASME graphite core components rules predicting probability of failure. Using the ASTM procedures yields failure curves at lower applied forces than the ASME rules. A journal paper was published in the Journal of Nuclear Engineering and Design exploring the statistical models of fracture in graphite.

Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

Energy Technology Data Exchange (ETDEWEB)

Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

2015-11-15

Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

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

Energy Technology Data Exchange (ETDEWEB)

1977-09-01

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

Non-catalyzed cathodic oxygen reduction at graphite granules in microbial fuel cells

International Nuclear Information System (INIS)

Freguia, Stefano; Rabaey, Korneel; Yuan Zhiguo; Keller, Juerg

2007-01-01

Oxygen is the most sustainable electron acceptor currently available for microbial fuel cell (MFC) cathodes. However, its high overpotential for reduction to water limits the current that can be produced. Several materials and catalysts have previously been investigated in order to facilitate oxygen reduction at the cathode surface. This study shows that significant stable currents can be delivered by using a non-catalyzed cathode made of granular graphite. Power outputs up to 21 W m -3 (cathode total volume) or 50 W m -3 (cathode liquid volume) were attained in a continuous MFC fed with acetate. These values are higher than those obtained in several other studies using catalyzed graphite in various forms. The presence of nanoscale pores on granular graphite provides a high surface area for oxygen reduction. The current generated with this cathode can sustain an anodic volume specific COD removal rate of 1.46 kg COD m -3 d -1 , which is higher than that of a conventional aerobic process. This study demonstrates that microbial fuel cells can be operated efficiently using high surface graphite as cathode material. This implies that research on microbial fuel cell cathodes should not only focus on catalysts, but also on high surface area materials

Non-catalyzed cathodic oxygen reduction at graphite granules in microbial fuel cells

Energy Technology Data Exchange (ETDEWEB)

Freguia, Stefano; Rabaey, Korneel; Yuan, Zhiguo; Keller, Juerg [The University of Queensland, St. Lucia, Qld (Australia). Advanced Wastewater Management Centre

2007-12-01

Oxygen is the most sustainable electron acceptor currently available for microbial fuel cell (MFC) cathodes. However, its high overpotential for reduction to water limits the current that can be produced. Several materials and catalysts have previously been investigated in order to facilitate oxygen reduction at the cathode surface. This study shows that significant stable currents can be delivered by using a non-catalyzed cathode made of granular graphite. Power outputs up to 21 W m{sup -3} (cathode total volume) or 50 W m{sup -3} (cathode liquid volume) were attained in a continuous MFC fed with acetate. These values are higher than those obtained in several other studies using catalyzed graphite in various forms. The presence of nanoscale pores on granular graphite provides a high surface area for oxygen reduction. The current generated with this cathode can sustain an anodic volume specific COD removal rate of 1.46 kg{sub COD} m{sup -3} d{sup -1}, which is higher than that of a conventional aerobic process. This study demonstrates that microbial fuel cells can be operated efficiently using high surface graphite as cathode material. This implies that research on microbial fuel cell cathodes should not only focus on catalysts, but also on high surface area materials. (author)

The investigation of HTGR fuel regeneration process

Energy Technology Data Exchange (ETDEWEB)

Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

1985-07-01

The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

The investigation of HTGR fuel regeneration process

International Nuclear Information System (INIS)

Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

1985-01-01

The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

International Nuclear Information System (INIS)

Tattersall, R.B.; Small, V.G.; MacBean, I.J.; Howe, W.D.

1964-08-01

The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

Nerva Fuel Element Development Program Summary Report - July 1966 through June 1972 Extrusion Studies

Energy Technology Data Exchange (ETDEWEB)

Napier, J. M.

1973-09-21

This part of the completion report pertaining to the NERVA graphite fuel element program covers data collected during the extrusion studies. The physical properties of the fuel element reached the following values: coefficient of thermal expansion (CTE) - 7.0 x 10-6/o C (25 - l,OOOo C); modulus of elasticity - 1.5 x lo6 psi; flexural strength - - 8,000 psi; ultimate strain to failure - 5,500 pidin; good thermal stress resistance. Matrices were produced which could be vapor coated with crack-free films of zirconium carbide. The CTE of the matrix was almost equal to the CTE of the zirconium carbide coating.

Estimation of graphite dust production in ITER TBM using finite element method

Energy Technology Data Exchange (ETDEWEB)

Kang, Ji-Ho, E-mail: jhkang@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daekeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Eung Seon [Korea Atomic Energy Research Institute, 989-111, Daekeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Ahn, Mu-Young; Lee, Youngmin; Park, Yi-Hyun; Cho, Seungyon [National Fusion Research Institute, 169-148, Gwahak-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2015-12-15

Highlights: • Graphite dust production was estimated for the Korean Helium Cooled Ceramic Reflector. • Wear amount was calculated by Archard model using finite element analysis results. • Life time estimation of graphite dust production was done. - Abstract: In this study, an estimation method of graphite dust production in the pebble-bed type reflector region of the Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER) project using Finite Element Method (FEM) was proposed and the total amount of dust production was calculated. A unit-cell model of uniformly arranged pebbles was defined with thermal and mechanical loadings. A commercial FEM program, Abaqus V6.10, was used to model and solve the stress field under multiple contact constraints between pebbles in the unit-cell. Resultant normal contact forces and slip distances on the contact points were applied into the Archard adhesive wear model to calculate the amount of graphite dust. The Finite Element (FE) analysis was repeated at 27 unit-cell locations chosen to form an interpolated dust density function for the entire region of the reflector. The dust production calculation was extended to the life time of the HCCR and the total graphite dust production was estimated to 0.279 g at the end of the life time with the maximum graphite dust density of 0.149 μg/mm{sup 3}. The dust explosion could be a safety issue with the calculated dust density level and it requires that an appropriate maintenance to remove sufficient amount of graphite dust regularly to prevent the possibility of dust explosion.

Estimation of graphite dust production in ITER TBM using finite element method

International Nuclear Information System (INIS)

Kang, Ji-Ho; Kim, Eung Seon; Ahn, Mu-Young; Lee, Youngmin; Park, Yi-Hyun; Cho, Seungyon

2015-01-01

Highlights: • Graphite dust production was estimated for the Korean Helium Cooled Ceramic Reflector. • Wear amount was calculated by Archard model using finite element analysis results. • Life time estimation of graphite dust production was done. - Abstract: In this study, an estimation method of graphite dust production in the pebble-bed type reflector region of the Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER) project using Finite Element Method (FEM) was proposed and the total amount of dust production was calculated. A unit-cell model of uniformly arranged pebbles was defined with thermal and mechanical loadings. A commercial FEM program, Abaqus V6.10, was used to model and solve the stress field under multiple contact constraints between pebbles in the unit-cell. Resultant normal contact forces and slip distances on the contact points were applied into the Archard adhesive wear model to calculate the amount of graphite dust. The Finite Element (FE) analysis was repeated at 27 unit-cell locations chosen to form an interpolated dust density function for the entire region of the reflector. The dust production calculation was extended to the life time of the HCCR and the total graphite dust production was estimated to 0.279 g at the end of the life time with the maximum graphite dust density of 0.149 μg/mm"3. The dust explosion could be a safety issue with the calculated dust density level and it requires that an appropriate maintenance to remove sufficient amount of graphite dust regularly to prevent the possibility of dust explosion.

Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

International Nuclear Information System (INIS)

Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R.

2015-01-01

It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

Prediction of the thermal behavior of a particle spherical fuel element using GITT

International Nuclear Information System (INIS)

Pessoa, C.V.; Oliveira, Claudio L. de; Jian, Su

2008-01-01

In this work, the transient and steady state heat conduction in a spherical fuel element of a pebble-bed high temperature were studied. This pebble element is composed by a particulate region with spherical inclusions, the fuel UO 2 particles, dispersed in a graphite matrix. A convective heat transfer by helium occurs on the outer surface of the fuel element. The two-energy equation model for the case of pure conduction was applied to this particulate spherical element, generating two macroscopic temperatures, respectively, of the inclusions and of the matrix. The transient analysis was carried out by using the Generalized Integral Transform Technique (GITT) that requires low computational efforts and allows a fast evaluation of the two macroscopic transient temperatures of the particulate region. The solution by GITT leads to a system of ordinary differential equations with the unknown transformed potentials. The mechanical properties (thermal conductivity and specific heat) of the materials were supposed not to depend on the temperature and to be uniform in each region. (author)

Principle design and data of graphite components

International Nuclear Information System (INIS)

Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

2004-01-01

The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

Prediction calculation of HTR-10 fuel loading for the first criticality

International Nuclear Information System (INIS)

Jing Xingqing; Yang Yongwei; Gu Yuxiang; Shan Wenzhi

2001-01-01

The 10 MW high temperature gas cooled reactor (HTR-10) was built at Institute of Nuclear Energy Technology, Tsinghua University, and the first criticality was attained in Dec. 2000. The high temperature gas cooled reactor physics simulation code VSOP was used for the prediction of the fuel loading for HTR-10 first criticality. The number of fuel element and graphite element was predicted to provide reference for the first criticality experiment. The prediction calculations toke into account the factors including the double heterogeneity of the fuel element, buckling feedback for the spectrum calculation, the effect of the mixture of the graphite and the fuel element, and the correction of the diffusion coefficients near the upper cavity based on the transport theory. The effects of impurities in the fuel and the graphite element in the core and those in the reflector graphite on the reactivity of the reactor were considered in detail. The first criticality experiment showed that the predicted values and the experiment results were in good agreement with little relative error less than 1%, which means the prediction was successful

Nuclear fuel element

International Nuclear Information System (INIS)

Penrose, R.T.; Thompson, J.R.

1976-01-01

A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

Fuel element loading system

International Nuclear Information System (INIS)

Arya, S.P; s.

1978-01-01

A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

International Nuclear Information System (INIS)

Baxter, A.M.; McEachern, D.; Hanson, D.L.; Vollman, R.E.

1994-11-01

The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted

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

Energy Technology Data Exchange (ETDEWEB)

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

1971-11-15

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

Graphite fuels combustion off-gas treatment options

International Nuclear Information System (INIS)

Kirkham, R.J.; Lords, R.E.

1993-03-01

Scenarios for burning bulk graphite and for burning crushed fuel particles from graphite spent nuclear fuels have been considered. Particulates can be removed with sintered metal filters. Subsequent cooling would then condense semi-volatile fission products into or onto a particulate. These particulates would be trapped by a second sintered metal filter or downstream packed bed. A packed bed scrub column can be used to eliminate most of the iodine-129 and tritium. A molecular sieve bed is proposed to collect the residual 129 I and other tramp radionuclides downstream (Ruthenium, etc.). Krypton-85 can be recovered, if need be, either by cryogenics or by the KALC process (Krypton Adsorption in Liquid Carbon dioxide). Likewise carbon-14 in the form of carbon dioxide could be collected with a caustic or lime scrub solution and incorporated into a grout. Sulfur dioxide present will be well below regulatory concern level of 4.0 tons per year and most of it would be removed by the scrubber. Carbon monoxide emissions will depend on the choice of burner and start-up conditions. Should the system exceed the regulatory concern level, a catalytic converter in the final packed bed will be provided. Radon and its daughters have sufficiently short half-lives (less than two minutes). If necessary, an additional holdup bed can be added before the final HEPA filters or additional volume can be added to the molecular sieve bed to limit radon emissions. The calculated total effective dose equivalent at the Idaho National Engineering Laboratory boundary from a single release of all the 3 , 14 C, 85 Kr, and 129 I in the total fuel mass if 0.43 mrem/year

Chapter 8: Exponential experiments on graphite moderated lattices fuelled by natural uranium tubes containing cylindrical graphite cores

International Nuclear Information System (INIS)

McCulloch, D.B.; Hoskins, T.A.

1963-01-01

Experiments have been carried out using a fuel element comprising a 2.75 in. o.d./2.40 in. i.d. natural uranium tube containing a graphite core of diameter 2.0 in. Values of material buckling and migration area asymmetry for lattices at 7 in., 8 in. and 8/2 in. pitch have been obtained, and correlated with the theory of Syrett (1961) to derive an effective resonance integral for the cored element. By comparison with the resonance integral for the same fuel tube without a core, a value for the constant 'γ' of the theory of Stace (1959) is obtained. (author)

Nuclear fuel element

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

Porous Carbon Materials for Elements in Low-Temperature Fuel Cells

Directory of Open Access Journals (Sweden)

Wlodarczyk R.

2015-04-01

Full Text Available The porosity, distribution of pores, shape of pores and specific surface area of carbon materials were investigated. The study of sintered graphite and commercial carbon materials used in low-temperature fuel cells (Graphite Grade FU, Toray Teflon Treated was compared. The study covered measurements of density, microstructural examinations and wettability (contact angle of carbon materials. The main criterion adopted for choosing a particular material for components of fuel cells is their corrosion resistance under operating conditions of hydrogen fuel cells. In order to determine resistance to corrosion in the environment of operation of fuel cells, potentiokinetic curves were registered for synthetic solution 0.1M H2SO4+ 2 ppmF-at 80°C.

Nuclear fuel elements

International Nuclear Information System (INIS)

Nakai, Keiichi

1983-01-01

Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

Nuclear fuel element

International Nuclear Information System (INIS)

Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

1980-01-01

Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

Calculation of thermal stresses in graphite fuel blocks

International Nuclear Information System (INIS)

Lejeail, Y.; Cabrillat, M.T.

2005-01-01

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

THE EFFECT OF GROUP IIIA TO VIA ELEMENTS AND THEIR OXIDES ON GRAPHITE OXIDATION

Energy Technology Data Exchange (ETDEWEB)

Rakszawski, J F; Parker, W E

1963-06-15

The effect of group IIIA to VIA elements and oxides on graphite oxidation was determined. Additives were molded with spectroscopically pure graphite powder. The concentration was maintained constant at 0.1 mole percent based on the element. The rate of reaction with 1 atm of air was measured at 700 and 800 deg C. Air flow rate from 2000 to 3000 cc/min had no effect on the oxidation rate of the pure graphite at 700, 750, and 800 deg C indicating that reaction was not occurring in Zone III. The calculated Ea of 54 kcal/mole suggested reaction in Zone I. Visual inspection of the rods after reaction substantiated this conclusion. The reaction was first order with respect to oxygen partial pressure at 700 and 800 deg C. B, B/sub 2/O/sub 5/, P, and P/sub 2/ O/sub 6/ inhibited the oxid ation of graphite at 700 and 800 deg C while the other elements and oxides catalyzed the reaction to various degrees. The reaction remained kinetically of the first order when inhibited. A systematic variation in reaction rates appears to follow the diagonals of the periodic relationship of the element from the upper left to the lower right. These variations can be correlated with average ionization energy or electron affinity. (auth)

Nuclear fuel elements design, fabrication and performance

CERN Document Server

Frost, Brian R T

1982-01-01

Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels

International Nuclear Information System (INIS)

Almeida, Eduardo S.; Silva, Luiz A.J.; Sousa, Raquel M.F.; Richter, Eduardo M.; Foster, Christopher W.; Banks, Craig E.; Munoz, Rodrigo A.A.

2016-01-01

This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L −1 HClO 4 (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. - Highlights: • Organic-resistant screen-printed graphitic electrodes (SPGE) for (bio)fuels. • Screen-printing of conductive and insulator inks on thin polyester substrate. • Continuous detection of antioxidants in electrolyte with 99% v/v ethanol. • SPGE coupled with batch-injection analysis allows over 200 injections (100 μL). • Similar results to GC and HPLC analyses of biodiesel and aviation jet fuels.

Fluid pressure method for recovering fuel pellets from nuclear fuel elements

International Nuclear Information System (INIS)

John, C.D. Jr.

1979-01-01

A method is described for removing fuel pellets from a nuclear fuel element without damaging the fuel pellets or fuel element sheath so that both may be reused. The method comprises holding the fuel element while a high pressure stream internally pressurizes the fuel element to expand the fuel element sheath away from the fuel pellets therein so that the fuel pellets may be easily removed

Fuel element

International Nuclear Information System (INIS)

Armijo, J.S.

1976-01-01

A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

(Fuel, fission product, and graphite technology)

Energy Technology Data Exchange (ETDEWEB)

Stansfield, O.M.

1990-07-25

Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

Hydrogen in CANDU fuel elements

International Nuclear Information System (INIS)

Sejnoha, R.; Manzer, A.M.; Surette, B.A.

1995-01-01

Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

Fluidized combustion of beds of large, dense particles in reprocessing HTGR fuel

International Nuclear Information System (INIS)

Young, D.T.

1977-03-01

Fluidized bed combustion of graphite fuel elements and carbon external to fuel particles is required in reprocessing high-temperature gas-cooled reactor (HTGR) cores for recovery of uranium. This burning process requires combustion of beds containing both large particles and very dense particles as well as combustion of fine graphite particles which elutriate from the bed. Equipment must be designed for optimum simplicity and reliability as ultimate operation will occur in a limited access ''hot cell'' environment. Results reported in this paper indicate that successful long-term operation of fuel element burning with complete combustion of all graphite fines leading to a fuel particle product containing <1% external carbon can be performed on equipment developed in this program

Nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Nuclear fuel element

International Nuclear Information System (INIS)

Thompson, J.R.; Rowland, T.C.

1976-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Technique for mass-spectrometric determination of moisture content in fuel elements and fuel element claddings

International Nuclear Information System (INIS)

Kurillovich, A.N.; Pimonov, Yu.I.; Biryukov, A.S.

1988-01-01

A technique for mass-spectroimetric determination of moisture content in fuel elements and fuek claddings in the 2x10 -4 -1.5x10 -2 g range is developed. The relative standard deviation is 0.13. A character of moisture extraction from oxide uranium fuels in the 20-700 deg C temperature range is studied. Approximately 80% of moisture is extracted from the fuels at 300 deg C. The moisture content in fuel elements with granular uranium oxide fuels is measured. Dependence of fuel element moisture content on conditions of hot vacuum drying is shown. The technique permits to optimize the fuel element fabrication process to decrease the moisture content in them. 4 refs.; 3 figs.; 2 tabs

Rack for nuclear fuel elements

International Nuclear Information System (INIS)

Rubinstein, H.J.; Gordon, C.B.; Robison, A.; Clark, P.M.

1977-01-01

Disclosed is a rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame. 16 claims, 14 figures

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Investigation on wear behavior of graphite baII under different pneumatic conveying environments

International Nuclear Information System (INIS)

Chen Zhipeng; Zheng Yanhua; Shi Lei; Yu Suyuan

2014-01-01

An experimental platform was built in the Institute of Nuclear and New Energy Technology (INET) to investigate the wear behavior of the graphite ball under the operational condition of the high temperature gas-cooled reactor (HTGR) fuel handling system. In this experimental platform, a series of experiments were carried out under different pneumatic conveying environments with the graphite balls, which were made of the material same as the fuel element matrix graphite (A3) of the 10 MW high temperature gas cooled reactor (HTR-10). The effect of the pneumatic conveying condition on the wear rate of graphite ball has been investigated, and the results include: (1) There is an obvious linear relationship between the wear rate and the feeding velocity of graphite ball elevated in the stainless steel elevating tube, and the wear rate will increase with the increase of the feeding velocity. (2) The wear rate of graphite ball under helium environment is significantly greater than that under air and nitrogen environments, which is caused by the different effects of various gas environments on mechanical properties of graphite. (author)

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

International Nuclear Information System (INIS)

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

1992-02-01

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

Fuel element

International Nuclear Information System (INIS)

Kennedy, S.T.

1982-01-01

A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

International Nuclear Information System (INIS)

Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

2009-10-01

This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

Nuclear fuel element

International Nuclear Information System (INIS)

Mogard, J.H.

1977-01-01

A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels

Energy Technology Data Exchange (ETDEWEB)

Almeida, Eduardo S.; Silva, Luiz A.J.; Sousa, Raquel M.F.; Richter, Eduardo M. [Universidade Federal de Uberlândia, Universidade Federal de Uberlândia, Av. João Naves de Ávila, 2121, Uberlândia, MG, 38408100 (Brazil); Foster, Christopher W.; Banks, Craig E. [Manchester Metropolitan University, Faculty of Science and the Environment, School of Science and the Environment, Division of Chemistry and Environmental Science, Manchester, M1 5GD, England (United Kingdom); Munoz, Rodrigo A.A., E-mail: raamunoz@iqufu.ufu.br [Universidade Federal de Uberlândia, Universidade Federal de Uberlândia, Av. João Naves de Ávila, 2121, Uberlândia, MG, 38408100 (Brazil)

2016-08-31

This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L{sup −1} HClO{sub 4} (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. - Highlights: • Organic-resistant screen-printed graphitic electrodes (SPGE) for (bio)fuels. • Screen-printing of conductive and insulator inks on thin polyester substrate. • Continuous detection of antioxidants in electrolyte with 99% v/v ethanol. • SPGE coupled with batch-injection analysis allows over 200 injections (100 μL). • Similar results to GC and HPLC analyses of biodiesel and aviation jet fuels.

Analytical study of stress and deformation of HTR fuel blocks

International Nuclear Information System (INIS)

Tanaka, M.

1982-01-01

A two-dimensional finite element computer code named HANS-GR has been developed to predict the mechanical behavior of the graphite fuel blocks with realistic material properties and core environment. When graphite material is exposed to high temperature and fast neutron flux of high density, strains arise due to thermal expansion, irradiation-induced shrinkage and creep. Thus stresses and distortions are induced in the fuel block in which there are spatial variation of these strains. The analytical method used in the program to predcit these induced stresses and distortions by finite element method is discussed. In order to illustrate the versatility of the computer code, numerical results of two example analyses of the multi-hole type fuel elements in the VHTR Reactor are given. Two example analyses presented are those concerning the stresses in fuel blocks with control rod holes and distortions of the fuel blocks at the periphery of the reactor core. It is considered these phenomena should be carefully examined when the multi-hole type fuel elements are applied to VHTR. It is assured that the predicted mechanical behavior of the graphite components is strongly dependent on the material properties used and obtaining the reliable material property is important to make the analytical prediction a reliable one

Monitoring arrangement for vented nuclear fuel elements

International Nuclear Information System (INIS)

Campana, R.J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180 0 rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements

Nuclear fuel elements

International Nuclear Information System (INIS)

Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.

1980-01-01

Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

Analysis of Off Gas From Disintegration Process of Graphite Matrix by Electrochemical Method

International Nuclear Information System (INIS)

Tian Lifang; Wen Mingfen; Chen Jing

2010-01-01

Using electrochemical method with salt solutions as electrolyte, some gaseous substances (off gas) would be generated during the disintegration of graphite from high-temperature gas-cooled reactor fuel elements. The off gas is determined to be composed of H 2 , O 2 , N 2 , CO 2 and NO x by gas chromatography. Only about 1.5% graphite matrix is oxidized to CO 2 . Compared to the direct burning-graphite method, less off gas,especially CO 2 , is generated in the disintegration process of graphite by electrochemical method and the treatment of off gas becomes much easier. (authors)

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1993-01-01

When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1994-01-01

When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

Energy Technology Data Exchange (ETDEWEB)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho; Lee, Young-woo; Cho, Moon-sung

2015-01-01

High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

Nuclear fuel element leak detection system

International Nuclear Information System (INIS)

John, C.D. Jr.

1978-01-01

Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

Method of producing exfoliated graphite composite compositions for fuel cell flow field plates

Energy Technology Data Exchange (ETDEWEB)

Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

2014-04-08

A method of producing an electrically conductive composite composition, which is particularly useful for fuel cell bipolar plate applications. The method comprises: (a) providing a supply of expandable graphite powder; (b) providing a supply of a non-expandable powder component comprising a binder or matrix material; (c) blending the expandable graphite with the non-expandable powder component to form a powder mixture wherein the non-expandable powder component is in the amount of between 3% and 60% by weight based on the total weight of the powder mixture; (d) exposing the powder mixture to a temperature sufficient for exfoliating the expandable graphite to obtain a compressible mixture comprising expanded graphite worms and the non-expandable component; (e) compressing the compressible mixture at a pressure within the range of from about 5 psi to about 50,000 psi in predetermined directions into predetermined forms of cohered graphite composite compact; and (f) treating the so-formed cohered graphite composite to activate the binder or matrix material thereby promoting adhesion within the compact to produce the desired composite composition. Preferably, the non-expandable powder component further comprises an isotropy-promoting agent such as non-expandable graphite particles. Further preferably, step (e) comprises compressing the mixture in at least two directions. The method leads to composite plates with exceptionally high thickness-direction electrical conductivity.

The fuel element of the first charge for EL 4; presentation, main problems arising in the research, production problems; L'element combustible du 1. jeu de EL 4; presentation, problemes essentiels poses par l'etude, problemes de fabrication

Energy Technology Data Exchange (ETDEWEB)

Ringot, C; Bailly, H; Bujas, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The fuel element making up the first charge for EL-4 is made of slightly enriched uranium oxide canned in stainless steel. This fuel element makes it possible to operate the reactor in the safest conditions awaiting the development of the fuel which will be finally adopted; this will have a low absorption can: beryllium, or a zirconium copper alloy. The 500 mm assembly is made up of 19 small rods placed on 3 rings, inside a graphite jacket. The solution adopted was a solution using completely independent small rods. This report deals with possible problems resulting from their study and production. (authors) [French] L'element combustible du 1er jeu EL-4 est un element combustible a oxyde d'uranium legerement enrichi gaine d'acier inoxydable. C'est un element combustible permettant de faire fonctionner le reacteur EL 4 dans des conditions aussi sures que possible avant de mettre au point le combustible definitif qui sera a gaine peu absorbante: beryllium, ou alliage zirconium-cuivre. L'assemblage de longueur 500 mm est constitue de 19 crayons places sur 3 couronnes, a l'interieur d'une chemise de graphite. La solution adoptee a ete une solution a crayons independants les uns des autres. Ce rapport traite des problemes eventuels poses par leur etude et leur fabrication. (auteurs)

Measurements of bundle end flux peaking effects in 37-element CANDU PHW fuel

International Nuclear Information System (INIS)

French, P.M.

1977-10-01

Thermal neutron bundle end flux peaking factors have been measured in fresh 37-element Bruce reactor natural UO 2 clusters in heavy water moderator, both with and without staggered plenums at the fuel stack ends, in representative elements throughout the clusters. The measurements were made at a square lattice pitch of 28.58 cm with heavy water coolant. The results indicate that outer element peaking factors are 1.142 +- 0.009 for bundles containing no plenums, and 1.155 +- 0.006 and 1.177 +- 0.006 at the non-plenum and plenum element ends respectively, for bundles containing staggered plenums, irrespective of the azimuthal orientation between pairs of bundles. Measurements are also reported for bundles containing plenums in every outer element, for bundles separated by a stainless steel flux suppressor, for longer graphite plenums, and for changes in plenum and bundle gap lengths. Some theoretical comparisons with the results, reported by other authors, have been summarized. (author)

Destruction of nuclear graphite using closed chamber incineration

International Nuclear Information System (INIS)

Senor, D.J.; Hollenberg, G.W.; Morgan, W.C.; Marianowski, L.G.

1994-01-01

Closed chamber incineration (CCI) is a novel technique by which irradiated nuclear graphite may be destroyed without the risk of radioactive cation release into the environment. The process utilizes an enclosed combustion chamber coupled with molten carbonate fuel cells (MCFCs). The transport of cations is intrinsically suppressed by the MCFCs, such that only the combustion gases are conducted through for release to the environment. An example CCI design was developed which had as its goal the destruction of graphite fuel elements from the Fort St. Vrain reactor (FSVR). By employing CCI, the volume of high level waste from the FSVR will be reduced by approximately 87 percent. Additionally, the incineration process will convert the SiC coating on the FSVR fuel particles to SiO 2 , thus creating a form potentially suitable for direct incorporation in a vitrification process stream. The design is compact, efficient, and makes use of currently available technology

Spacer for supporting fuel element boxes

International Nuclear Information System (INIS)

Wild, E.

1979-01-01

A spacer plate unit arranged externally on each side and at a predetermined level of a polygonal fuel element box for mutually supporting, with respect to one another, a plurality of the fuel element boxes forming a fuel element bundle, is formed of a first and a second spacer plate part each having the same length and the same width and being constituted of unlike first and second materials, respectively. The first and second spacer plate parts of the several spacer plate units situated at the predetermined level are arranged in an alternating continuous series when viewed in the peripheral direction of the fuel element box, so that any two spacer plate units belonging to face-to-face oriented sides of two adjoining fuel element boxes in the fuel element bundle define interfaces of unlike materials

Unified fuel elements development for research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1998-01-01

Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

Automated Fuel Element Closure Welding System

International Nuclear Information System (INIS)

Wahlquist, D.R.

1993-01-01

The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

Analysis of electrochemical disintegration process of graphite matrix

International Nuclear Information System (INIS)

Tian Lifang; Wen Mingfen; Chen Jing

2010-01-01

The electrochemical method with ammonium nitrate as electrolyte was studied to disintegrate the graphite matrix from the simulative fuel elements for high temperature gas-cooled reactor. The influences of process parameters, including salt concentration, system temperature and current density, on the disintegration rate of graphite fragments were investigated in the present work. The experimental results showed that the disintegration rate depended slightly on the temperature and salt concentration. The current density strongly affected the disintegration rate of graphite fragments. Furthermore, the content of introduced oxygen in final graphite fragments was independent of the current density and the concentration of electrolyte. Moreover, the structural evolution of graphite was analyzed based on the microstructural parameters determined by X-ray diffraction profile fitting analysis using MAUD (material analysis using diffraction) before and after the disintegration process. It may safely be concluded that the graphite disintegration can be ascribed to the influences of the intercalation of foreign molecules in between crystal planes and the partial oxidation involved. The disintegration process was described deeply composed of intercalate part and further oxidation part of carbon which effected together to lead to the collapse of graphite crystals.

Strategy Study on Treatment and Disposal of the Radioactive Graphite Waste of HTR

International Nuclear Information System (INIS)

Li Junfeng; Ma Tao; Wang Jianlong

2014-01-01

The possible options to change HTGR spent fuel into an acceptable form for repository disposal were discussed. The progresses of physical, chemical, and electrochemical separation of graphite from the HTGR fuel elements were reviewed. The advantages and disadvantages of each method were listed out. The total waste volume of each method was compared. The preferred option depends on the waste acceptance criteria for the repository, availability of low level waste disposal for graphite, overall economics, and overall risks. The minimum processing that yields an acceptable waste form also gives the lowest costs as well as the simplest process and the least risk. The options that could be used for treating HTGR spent fuels were listed out. The strategy for treating HTGR spent fuels and the packages needed for repository were discussed. (author)

HTGR fuel cycle

International Nuclear Information System (INIS)

1987-08-01

In the spring of 1987, the HTGR fuel cycle project has been existing for ten years, and for this reason a status seminar has been held on May 12, 1987 in the Juelich Nuclear Research Center, that gathered the participants in this project for a discussion on the state of the art in HTGR fuel element development, graphite development, and waste management. The papers present an overview of work performed so far and an outlook on future tasks and goals, and on taking stock one can say that the project has been very successful so far: The HTGR fuel element now available meets highest requirements and forms the basis of today's HTGR safety philosophy; research work on graphite behaviour in a high-temperature reactor has led to complete knowledge of the temperature or neutron-induced effects, and with the concept of direct ultimate waste disposal, the waste management problem has found a feasible solution. (orig./GL) [de

Nuclear reactor fuel element assemblies

International Nuclear Information System (INIS)

Krawiec, D.M.; Bevilacqua, F.

1974-01-01

The fuel elements of each fuel element group are separated from each other by means of a multitude of thin, intersecting plates in the from of grid strips. Flow deflectors near the surface of the fuel elements are used in order to make the coolant flow more turbulent. They are designed as vanes and arranged at a distance on the grid strips. Each deflector vane has two arms stretching in opposite directions, each one into a neighbouring channel. In outward direction, the deflector vanes are converging. The strips with the vanes can be put on the supporting grid of the fuel elements. The vane structure can be reinforced by providing distortions in the strip material near the vanes. (DG) [de

Graphitic Carbon Nitride as a Catalyst Support in Fuel Cells and Electrolyzers

International Nuclear Information System (INIS)

Mansor, Noramalina; Miller, Thomas S.; Dedigama, Ishanka; Jorge, Ana Belen; Jia, Jingjing; Brázdová, Veronika; Mattevi, Cecilia; Gibbs, Chris; Hodgson, David; Shearing, Paul R.; Howard, Christopher A.; Corà, Furio; Shaffer, Milo; Brett, Daniel J.L.

2016-01-01

Highlights: • Graphitic carbon nitride (gCN) describes many materials with different structures. • gCNs can exhibit excellent mechanical, chemical and thermal resistance. • A major obstacle for pure gCN catalyst supports is limited electronic conductivity. • Composite/Hybrid gCN structures show excellent performance as catalyst supports. • gCNs have great potential for use in fuel calls and water electrolyzers. - Abstract: Electrochemical power sources, such as polymer electrolyte membrane fuel cells (PEMFCs), require the use of precious metal catalysts which are deposited as nanoparticles onto supports in order to minimize their mass loading and therefore cost. State-of-the-art/commercial supports are based on forms of carbon black. However, carbon supports present disadvantages including corrosion in the operating fuel cell environment and loss of catalyst activity. Here we review recent work examining the potential of different varieties of graphitic carbon nitride (gCN) as catalyst supports, highlighting their likely benefits, as well as the challenges associated with their implementation. The performance of gCN and hybrid gCN-carbon materials as PEMFC electrodes is discussed, as well as their potential for use in alkaline systems and water electrolyzers. We illustrate the discussion with examples taken from our own recent studies.

Method of measuring distance between fuel element

International Nuclear Information System (INIS)

Urata, Megumu.

1991-01-01

The distance between fuel elements contained in a pool is measured in a contactless manner even for a narrow distance less than 1 mm. That is, the equipment for measuring the distance between spent fuel elements of a spent fuel assembly in a nuclear reactor comprises a optical fiber scope, a lens, an industrial TV camera and a monitor TV. The top end of the optical fiber scope is inserted between fuel elements to be measured. The state thereof is displayed on the TV screen to measure the distance between the fuel elements. The measured results are compared with a previously formed calibration curve to determine the value between the fuel elements. Then, the distance between the fuel elements can be determined in the pool of a power plant without dismantling the fuel assembly, to investigate the state of the bending and estimate the fuel working life. (I.S.)

Brazed graphite/refractory metal composites for first-wall protection elements

International Nuclear Information System (INIS)

Smid, I.; Croessmann, C.D.; Salmonson, J.C.; Whitley, J.B.; Nickel, H.

1991-01-01

The peak surface heat flux deposition on divertor elements of near term fusion devices is expected to exceed 10 MW/m 2 . The needed reliability of brazed plasma interactive components, particularly under abnormal operating conditions with peak surface temperatures well beyond 1000deg C, makes refractory metallic substrates and brazes with a high melting point very attractive. TZM, a high temperature alloy of molybdenum, and isotropic graphite, materials very closely matched in their thermal expansion, were brazed with four high-temperature brazes. The brazes used were Zr, 90Ni/10Ti, 90Cu/10Ti and 70Ag/27Cu/3Ti (nominal composition prior to brazing, wt%). The resulting composite tiles of 50x50 mm 2 with a TZM thickness of 5 mm and a graphite thickness of 10 mm have been tested in high heat flux simulation for their thermal fatigue properties. Up to 600 loading cycles were carried out with an average heat flux of 10 MW/m 2 for 0.5 s pulses. The maximum surface temperature was 1100deg C. In support of the experiment, the thermal response and temperature gradients of the samples were investigated using a finite element model. (orig.)

Brazed graphite/refractory metal composites for first-wall protection elements

International Nuclear Information System (INIS)

Smid, I.; Croessmann, C. D.; Salmonson, J. C.; Whitley, J. B.; Kny, E.; Reheis, N.; Kneringer, G.; Nickel, H.

1995-01-01

The peak surface heat flux deposition on divertor elements of near term fusion devices is expected to exceed 10 MW/m 2 . The needed reliability of brazed plasma interactive components, particularly under abnormal operating conditions with peak surface temperatures well beyond 1000 degree C, makes refractory metallic substrates and brazes with a high melting point very attractive. TZM, a high temperature alloy of molybdenum, and isotropic graphite, materials very closely matched in their thermal expansion, were brazed with four high-temperature brazes. The brazes used were Zr, 90Ni/10Ti, 90Cu/10Ti and 70Ag/27Cu/3Ti (nominal composition prior to brazing, wt%). The resulting composite tiles of 5O X 50 mm 2 with a TZM thickness of 5 mm and a graphite thickness of 10 mm have been tested in high heat flux simulation for their thermal fatigue properties. Up to 600 loading cycles were carried out with an average heat flux of 10 MW/m 2 for 0.5 s pulses. The maximum surface temperature was 1100 degree C. In support of the experiment, the thermal response and temperature gradients of the samples were investigated using a finite element model. (author)

Brazed graphite/refractory metal composites for first wall protection elements

International Nuclear Information System (INIS)

Smid, I.; Croessmann, C. D.; Salmonson, J. C.; Whitley, J. B.; Kny, E.; Reheis, N; Kneringer, G.; Nickel, H.

1995-01-01

The peak surface heat flux deposition on divertor elements of near term fusion devices is expected to exceed 10 MW/m 2 . The needed reliability of brazed plasma interactive components, particularly under abnormal operating conditions with peak surface temperatures well beyond 1000 degree C, makes refractory metallic substrates and brazes with a high melting point very attractive. TZM, a high temperature alloy of molybdenum, and isotropic graphite, materials very closely matched in their thermal expansion, were brazed with four high-temperature brazes. The brazes used were Zr, 90Ni/10Ti, 90Cu/10Ti and 70Ag/27Cu/10Ti (nominal composition prior to brazing, wt%). The resulting composite tiles of 50 x 50 mm with a TZM thickness of 5 mm and a graphite thickness of 10 mm have been tested in high heat flux simulation for their thermal fatigue properties. Up to 600 loading cycles were carried out with the experimental parameters chosen to cover NET/ITER design specifications. In support of the experiment, the thermal response and temperature gradients of the samples were investigated using a finite element model. (author)

Brazed graphite/refractory metal composites for first-wall protection elements

Science.gov (United States)

Šmid, I.; Croessmann, C. D.; Salmonson, J. C.; Whitley, J. B.; Kny, E.; Reheis, N.; Kneringer, G.; Nickel, H.

1991-03-01

The peak surface heat flux deposition on divertor elements of near term fusion devices is expected to exceed 10 MW/m 2. The needed reliability of brazed plasma interactive components, particularly under abnormal operating conditions with peak surface temperatures well beyond 1000°C, makes refractory metallic substrates and brazes with a high melting point very attractive. TZM, a high temperature alloy of molybdenum, and isotropic graphite, materials very closely matched in their thermal expansion, were brazed with four high-temperature brazes. The brazes used were Zr, 90Ni/10Ti, 90Cu/10Ti and 70Ag/27Cu/3Ti (nominal composition prior to brazing, wt%). The resulting composite tiles of 50 × 50 mm2 with a TZM thickness of 5 mm and a graphite thickness of 10 mm have been tested in high heat flux simulation for their thermal fatigue properties. Up to 600 loading cycles were carried out with an average heat flux of 10 MW/m 2 for 0.5 s pulses. The maximum surface temperature was 1100°C. In support of the experiment, the thermal response and temperature gradients of the samples were investigated using a finite element model.

Increased burnup of fuel elements

International Nuclear Information System (INIS)

Ahlf, J.

1983-01-01

The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.) [de

Pre-irradiation testing of experimental fuel elements

International Nuclear Information System (INIS)

Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

1979-01-01

The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

Fuel element transport container

International Nuclear Information System (INIS)

Benna, P.; Neuenfeldt, W.

1979-01-01

The reprocessing system includes a large number of waterfilled ponds next to each other for the intermediate storage of fuel elements from LWR's. The fuel element transport device is allocated to a middle pond. The individual ponds are separated from each other by walls, and are only accessible from the middle pond via narrow passages. The transport device includes a telescopic running rail for a trolley with a grab device for the fuel element. The running rail is supported in turn by a second trolley, which can be moved by wheels on rails. Part of the drive of the first trolley is arranged on the second one. Using this transport device, adjacent ponds can be served through the passage openings. (DG) [de

The problem of gas gap between graphite - fuel channel reduction impact at Ignalina NPP

International Nuclear Information System (INIS)

1999-01-01

Safety analysis of Ignalina NPP operation in the case when gap closure between graphite - fuel channel occur was performed. The main results of this analysis as well as data of gap measurements during the year 1996 - 1998 are provided

Laminated exfoliated graphite composite-metal compositions for fuel cell flow field plate or bipolar plate applications

Science.gov (United States)

Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

2014-05-20

An electrically conductive laminate composition for fuel cell flow field plate or bipolar plate applications. The laminate composition comprises at least a thin metal sheet having two opposed exterior surfaces and a first exfoliated graphite composite sheet bonded to the first of the two exterior surfaces of the metal sheet wherein the exfoliated graphite composite sheet comprises: (a) expanded or exfoliated graphite and (b) a binder or matrix material to bond the expanded graphite for forming a cohered sheet, wherein the binder or matrix material is between 3% and 60% by weight based on the total weight of the first exfoliated graphite composite sheet. Preferably, the first exfoliated graphite composite sheet further comprises particles of non-expandable graphite or carbon in the amount of between 3% and 60% by weight based on the total weight of the non-expandable particles and the expanded graphite. Further preferably, the laminate comprises a second exfoliated graphite composite sheet bonded to the second surface of the metal sheet to form a three-layer laminate. Surface flow channels and other desired geometric features can be built onto the exterior surfaces of the laminate to form a flow field plate or bipolar plate. The resulting laminate has an exceptionally high thickness-direction conductivity and excellent resistance to gas permeation.

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydın; Kazimi, Mujid S.

2013-01-01

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

2013-10-15

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Experimental research of fuel element reliability

International Nuclear Information System (INIS)

Cech, B.; Novak, J.; Chamrad, B.

1980-01-01

The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

Critical experiments on enriched uranium graphite moderated cores

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

1978-07-01

A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

Fuel element tomography by gammametry

International Nuclear Information System (INIS)

Simonet, G.; Pineira, T.

1982-03-01

As from transversal gamma determinations of a cylindrical fuel element, the TOMOGAM program reconstitutes the distribution of fission products in a section. This direct, fast and non destructive method, makes it possible to have access to the behaviour of the fuel at any time: - the soluble fission products in the matrix represent the fuel itself and the distribution of the fissions, - the migrating elements inform on the temperature reached in accordance with the permitted powers, - the volatile nuclides build up in particular points where physical-chemical phenomena of fuel-cladding interaction are liable to corrode the latter. Hence, gamma spectrometry extends its possibilities of analysis relative to the performance of reactor elements [fr

Nuclear fuels and development of nuclear fuel elements

International Nuclear Information System (INIS)

Sundaram, C.V.; Mannan, S.L.

1989-01-01

Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

Nuclear reactor fuel element splitter

International Nuclear Information System (INIS)

Yeo, D.

1976-01-01

A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

Burnup measurements of leader fuel elements

International Nuclear Information System (INIS)

Henriquez, C; Navarro, G; Pereda, C

2000-01-01

Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

PEM fuel cells with injection moulded bipolar plates of highly filled graphite compounds; PEM-Brennstoffzellen mit spritzgegossenen Bipolarplatten aus hochgefuelltem Graphit-Compound

Energy Technology Data Exchange (ETDEWEB)

Kreuz, Can

2008-04-11

This work concerns with the injection moulding of highly filled graphite compounds to bipolar plates for PEM fuel cells in a power output range between 100 - 500 Watts. A particular focus is laid on the combination of the three multidisciplinary scopes like material development, production technology and component development / design. The results of the work are specified by the process-oriented characterisation of the developed and manufactured bipolar plates as well as their application in a functioning fuel cell. (orig.)

Device for manipulating a nuclear reactor fuel element in a fuel element pond containing water

International Nuclear Information System (INIS)

Jabsen, F.S.

1977-01-01

Using this device a fuel element can be manipulated inside a water filled storage pond for inspection purposes. A transport arrangement which is normally situated above such a pond is modified for this purpose. A crane bridge runs on rails on the upper edge of the pond. A type of trolley runs transversely to the direction of travel of the bridge between 2 wide flange supports forming the crane support. During movement this trolley moves a submerged combination of periscope and TV camera pendant from it at about half the pond height horizontally along the crane support. 2 vehicles move between these on 4 rollers each, on the under flanges of the crane support at spacings of about one fuel element length. A pendant arm of the same length as the periscope dips vertically into the pond from each vehicle. There is a bar of about fuel element length resting on the lower ends of both arms. The surface of a fuel element lying on this bar can be inspected through the periscope on longitudinal travel of the trolley. The bar with the fuel element can be rotated 90 0 downwards into a vertical position after removal of one or more rotating kingpins and release of a rope hanging on the end away from the kingpin. The rope is actuated by a winch on the crane support. The bar has vertical plates at both ends to hold the fuel element in its vertical position. (HP) [de

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Tanihiro, Yasunori; Sumita, Isao.

1970-01-01

An improved fuel element of the heat pipe type is disclosed in which the fuel element itself is given a heat pipe structure and filled with a coated particle fuel at the section thereof having a capillary tube construction, whereby the particular advantages of heat pipes and coated fuels are combined and utilized to enhance thermal control and reactor efficiency. In an embodiment, the fuel element of the present invention is filled at its lower capillary tube section with coated fuel and at its upper section with a granurated neutron absorber. Both sections are partitioned from the central shaft by a cylindrically shaped wire mesh defining a channel through which the working liquid is vaporized from below and condensed by the coolant external to the fuel element. If the wire mesh is chosen to have a melting point lower than that of the fuel but higher than that of the operating temperature of the heat pipe, the mesh will melt and release the neutron absorbing particles should hot spots develop, thus terminating fission. (Owens, K. J.)

Loads on pebble bed fuel elements

Energy Technology Data Exchange (ETDEWEB)

Teuchert, E.; Maly, V.

1974-03-15

A comparison is made of key parameters for multi-recycle pebbles and single-pass once-through (OTTO) pebbles. The parameters analyzed include heat transfer characteristics with burn-up, temperature profiles, power per element as a function of axial position in the core, and burn-up. For the OTTO-scheme, the comparisons addressed the use of the conventional fuel element and the advanced "shell ball" designed to reduce the peak fuel temperature in the center of the fuel element. All studies addressed the uranium-thorium fuel cycle.

Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

International Nuclear Information System (INIS)

Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A.

1999-01-01

The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

System for assembling nuclear fuel elements

International Nuclear Information System (INIS)

1980-01-01

An automatic system is described for assembling nuclear fuel elements, in particular those employing mixed oxide fuels. The system includes a sealing mechanism which allows movement during the assembling of the fuel element along the assembly stations without excessive release of contaminants. (U.K.)

Fuel Element Technical Manual

Energy Technology Data Exchange (ETDEWEB)

Burley, H.H. [ed.

1956-08-01

It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

Inert matrix fuel in dispersion type fuel elements

Energy Technology Data Exchange (ETDEWEB)

Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

2006-06-30

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Inert matrix fuel in dispersion type fuel elements

Science.gov (United States)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Development and testing of the EDF-2 reactor fuel element; Essais et mise au point de l'element combustible pour le reacteur EDF-2

Energy Technology Data Exchange (ETDEWEB)

Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Furhmann, R [Societe Industrielle de Combustible Nucleaire (France)

1964-07-01

This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the {beta} phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [French] Ce rapport technique

VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

Energy Technology Data Exchange (ETDEWEB)

Boehnert, R.; Walter, C.

1977-02-15

The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given. (LK)

VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

International Nuclear Information System (INIS)

Boehnert, R.; Walter, C.

The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given

Coherence of reactor design and fuel element design

International Nuclear Information System (INIS)

Vom Scheidt, S.

1995-01-01

Its background of more than 25 years of experience makes Framatome the world's leading company in the design and sales of fuel elements for pressurized water reactors (PWR). In 1994, the fuel fabrication units were incorporated as subsidiaries, which further strengthens the company's position. The activities in the fuel sector comprise fuel element design, selection and sourcing of materials, fuel element fabrication, and the services associated with nuclear fuel. Design responsibility lies with the Design and sales Management, which closely cooperates with the engineers of the reactor plant for which the fuel elements are being designed, for fuel elements are inseparable parts of the respective reactors. The Design and Sales Management also has developed a complete line of services associated with fuel element inspection and repair. As far as fuel element sales are concerned, Framatome delivers the first core in order to be able to assume full responsibility vis-a-vis the customer for the performance of the nuclear steam supply system. Reloads are sold through the Fragema Association established by Framatome and Cogema. (orig.) [de

Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels.

Science.gov (United States)

Almeida, Eduardo S; Silva, Luiz A J; Sousa, Raquel M F; Richter, Eduardo M; Foster, Christopher W; Banks, Craig E; Munoz, Rodrigo A A

2016-08-31

This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L(-1) HClO4 (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. Copyright © 2016 Elsevier B.V. All rights reserved.

Costs of head-end incineration with respect to Kr separation in the reprocessing of HTR fuel elements

Energy Technology Data Exchange (ETDEWEB)

Barnert-Wiemer, H.; Boehnert, R.

1976-07-15

The C-incinerations and the Kr-separations during head-end incineration in the reprocessing of HTR fuel elements are described. The costs for constructing an operating a head-end incineration of reprocessing capacities with 5,000 to 50,000 MW(e)-HTR power have been determined. The cost estimates are divided into investment and operating costs, further after the fraction of the N/sub 2/-content in the incineration exhaust gas, which strongly affects costs. It appears that, in the case of Kr-separation from the incineration exhaust gas, the investment costs as well as the operating costs of the head-end for N/sub 2/-containing exhaust gas are considerably greater than those for gas without N/sub 2/. The C-incineration of the graphite of the HTR fuel elements should therefore only be performed with influx gas that is free of N/sub 2/.

Nuclear fuel element end fitting

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

A typical embodiment of the invention has an array of sockets that are welded to the intersections of the plates that form the upper and lower end fittings of a nuclear reactor fuel element. The sockets, which are generally cylindrical in shape, are oriented in directions that enable the longitudinal axes of the sockets to align with the longitudinal axes of the fuel rods that are received in the respective sockets. Detents impressed in the surfaces of the sockets engage mating grooves that are formed in the ends of the fuel rods to provide for the structural integrity of the fuel element

Method for inspecting nuclear reactor fuel elements

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

Integral nuclear fuel element assembly

International Nuclear Information System (INIS)

Schluderberg, D. C.

1985-01-01

An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Fast breeder fuel element development

International Nuclear Information System (INIS)

Marth, W.; Muehling, G.

1983-08-01

This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

Unification of fuel elements for research reactors

International Nuclear Information System (INIS)

Vatulyn, A.V.; Stetskyi, Y.A.; Dobrikova, I.V.

1997-01-01

To the purpose of fuel elements unification the possibility of rod fuel assembly (FA) using in the cores of research reactors have been considered in this paper. The calculation results of geometric, hydraulic and thermotechnical parameters of rod assembly are submitted. Several designs of finned square fuel element and fuel assembly are proposed on base of analysis of rod FA characteristics in compare of tube ones. The fuel elements specimens and the model assembly are manufactured. The developed designs are the basis for further optimization after neutron-physical calculations of cores. (author)

Apparatus for locating defective nuclear fuel elements

International Nuclear Information System (INIS)

Lawrie, W.E.

1979-01-01

An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

Hydriding failure in water reactor fuel elements

International Nuclear Information System (INIS)

Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

1980-01-01

Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

Nuclear fuel elements

International Nuclear Information System (INIS)

Ainsworth, K.F.

1979-01-01

A nuclear fuel element is described having a cluster of nuclear fuel pins supported in parallel, spaced apart relationship by transverse cellular braces within coaxial, inner and outer sleeves, the inner sleeve being in at least two separate axial lengths, each of the transverse braces having a peripheral portion which is clamped peripherally between the ends of the axial lengths of the inner sleeve. (author)

LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

International Nuclear Information System (INIS)

Zidan, W.I.

2000-01-01

The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

Reproduction of the RA reactor fuel element fabrication

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

Storage device for a long nuclear reactor fuel element and/or a long nuclear reactor fuel element part

International Nuclear Information System (INIS)

Vogt, M.; Schoenwitz, H.P.; Dassbach, W.

1986-01-01

The storage device can be erected in a dry storage room for new fuel elements and also in a storage pond for irradiated fuel elements. It consists of shells, which are arranged vertically and which have a lid. A suspension for the fuel element is provided on the underside of the lid, which acts as a support against squashing or bending in case of vertical forces acting (earthquake). (DG) [de

Chilean fuel elements fabrication progress report

International Nuclear Information System (INIS)

Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

1993-01-01

Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

Developments in natural uranium - graphite reactors

International Nuclear Information System (INIS)

Bourgeois, J.

1964-01-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

Thermal insulation of fuel elements

International Nuclear Information System (INIS)

Dubrovcak, P.; Pec, V.; Pitonak, J.

1978-01-01

The claim of the invention concerns thermal insulation of fuel elements heated for measurement of uranium fuel physical properties. For this, layers of aluminium film and of glass fibre are wound onto the inner tube of the element cladding. The space between the inner and the outer tubes is evacuated and the tubes are spaced using spacer wires. (M.S.)

Gravity Effects on the Free Vibration of a Graphite Column

International Nuclear Information System (INIS)

Ki, Dong-Ok; Kim, Jong-Bum; Park, Keun-Bae; Lee, Won-Jae

2006-01-01

The gravity effects on the free vibration of a graphite column are studied. Graphite block is a key component of a HTGR (High Temperature Gas Cooled Reactor). The major core elements, such as the fuel blocks and neutron reflector blocks, of HTTR (High Temperature Test Reactor, Japan) and GT-MHR (Gas Turbine- Modular Helium Reactor, USA) consist of stacked hexagonal graphite blocks forming a group of columns. The vibration of the columns induced by earthquakes may lead to solid impacts between graphite blocks and structural integrity problems. The study of free vibration characteristics of the graphite block column is the first step in the core internal structure dynamic analysis. Gravity force bring a negative stiffness term to the transverse vibration analysis of heavy long column structures, and results in natural frequency reductions. Generally it is not considered in the not so tall structure cases, because the gravity term makes the analysis and design complicated. Therefore it is important to check whether the gravity effect is severe or not

Nuclear fuel elements

International Nuclear Information System (INIS)

Obara, Hiroshi.

1981-01-01

Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

Graphite

Science.gov (United States)

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

2017-12-19

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

International experience in conditioning spent fuel elements

International Nuclear Information System (INIS)

Ashton, P.

1991-04-01

The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP) [de

Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

Energy Technology Data Exchange (ETDEWEB)

Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

2003-02-01

This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

Fuel element box inspection device

International Nuclear Information System (INIS)

Ortmayer, R.M.; Pick, W.

1985-01-01

The invention concerns a device for inspecting the outer geometry of a long fuel element box by measuring the surface contours over its longitudinal crossection and along its length by sensors. These are kept in a sledge which can be moved along the fuel element guide in a slot guide. The measurement signals reach an evaluation device outside the longitudinal box. (orig./HP) [de

Computer simulation of fuel element performance

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, G I

1979-01-01

The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

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)

Transport of fission products in matrix and graphite

International Nuclear Information System (INIS)

Hoinkis, E.

1983-06-01

In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

From Core to Capture: Graphite Management by Gasification and Carbon Capture & Storage (CCS)

International Nuclear Information System (INIS)

Goodwin, J.; Bradbury, D.; Black, S.; Tomlinson, T.; Livesey, B.; Robinson, J.; Lindberg, M.; Newton, C.; Jones, A.; Wickham, A.

2016-01-01

Radioactive graphite waste arises principally from the moderators of graphite/gas-cooled reactors at the end of life of the reactors. Commercial power producing reactors (for example, Magnox, AGR and RBMK) have graphite moderators, each containing several thousand tonnes of graphite, with the UK having the largest inventory of over 90,000 tonnes. Additionally, there are smaller quantities of graphite arising from other sources such as fuel element components. The current long term strategy for management of reactor graphite in the UK is for these wastes to be conditioned for disposal followed by transfer to a geological disposal facility (GDF). With this baseline position, these wastes will account for about 30% of the ILW inventory in a GDF. As the volume of the graphite waste is so large, it is not currently economic to retrieve and process the graphite in advance of the availability of a geological disposal facility. Recent work by the NDA has ascribed a much smaller “incremental” volume of 2% due to graphite, calculated on the basis that the GDF has to be a certain size anyway in order to dissipate the decay heat from high level waste

Catalogue of fuel elements - 1. addendum October 1958

International Nuclear Information System (INIS)

Even, A.

1957-01-01

This document contains sheets presenting various characteristics of nuclear fuel elements which are distinguished with respect to their shape: cylinder bar, plate, tube. Each sheet comprises an indication of the atomic pile in which the fuel element is used, dimensions, cartridge data, data related to cooling, to combustion rate, and to fuel handling. A drawing of the fuel element is also given

Quality assurance of fuel elements

International Nuclear Information System (INIS)

Hoerber, J.

1980-01-01

The quality assurance activities for reactor fuel elements are based on a quality assurance system which implies the requirements resulting from the specifications, regulations of the authorities, national standards and international rules and regulations. The quality assurance related to production of reactor fuel will be shown for PWR fuel elements in all typical fabrication steps as conversion into UO 2 -powder, pelletizing, rodmanufacture and assembling. A wide range of destructive and nondestructive techniques is applied. Quality assurance is not only verified by testing techniques but also by process monitoring by means of parameter control in production and testing procedures. (RW)

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Butterfield, C.E.; Waite, E.

1982-01-01

A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

Fuel element transfer cask modelling using MCNP technique

International Nuclear Information System (INIS)

Rosli Darmawan

2009-01-01

Full text: After operating for more than 25 years, some of the Reaktor TRIGA PUSPATI (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement. (author)

Fuel Element Transfer Cask Modelling Using MCNP Technique

International Nuclear Information System (INIS)

Darmawan, Rosli; Topah, Budiman Naim

2010-01-01

After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

Thermally-induced bowing of CANDU fuel elements

International Nuclear Information System (INIS)

Suk, H.C.; Sim, K.S.; Park, J.H.; Park, G.S.

1995-01-01

Considering only the thermally-induced bending moments which are generated both within the sheath and between the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element, a generalized and explicit analytical formula for the thermally-induced bending is developed in this paper, based on the cases of 1) the bending of an empty tube treated by neglecting of the fuel/sheath mechanical interaction and 2) the fuel/sheath interaction due to the pellet and sheath temperature variations. In each of the cases, the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. Investigating the relative importance of the various parameters affecting fuel element bowing, the element bowing is found to be greatly affected with the variations of element length, sheath diameter, pellet/sheath mechanical interaction and neutron flux depression factors, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient, and sheath and pellet thermal conductivities. Also, the element bowing of the standard 37-element bundle and CANFLEX 43-element bundle for the use in CANDU-6 reactors was analyzed with the formula, which could help to demonstrate the integrity of the fuel. All the required input data for the analyses were generated in terms of the reactor operation conditions on the reactor physics, thermal hydraulics and fuel performance by using various CANDU computer codes. The analysis results indicate that the CANFLEX 43-element

Fuel elements for LWR power plants

International Nuclear Information System (INIS)

Roepenack, H.

1977-01-01

About five times more expensive than the fabrication of a fuel element is the enriched uranium contained therein; soon the monthly interest charges for the uranium value of a fuel element reload will account for five percent of the fabrication costs, and much more expensive than all this together can it be if reactor operation has to be interrupted because of damaged elements. Thus, quality assurance comes first. (orig.) [de

Determination of molybdenum, ruthenium, rhodium, and palladium in radioinactive simulated waste of the nuclear fuel cycle by solid sampling graphite furnace atomic absorption spectrometry (GFAAS)

International Nuclear Information System (INIS)

Schmiedel, G.; Mainka, E.; Ache, H.J.

1989-01-01

In relation with insoluble particles in the nuclear fuel cycle waste, the solid sampling GFAAS was used to determine molybdenum, ruthenium, rhodium, and palladium in such waste. Two methods for the direct determination of these elements are described. The samples must be handled in glove boxes or moreover in hot cells with a robot. The determination of the elements by the cup-in-tube technique needs a very sensitive balance (microbalance) for weighing in μg-range and the handling of this method is not practical in glove boxes and hot cells. An alternative technique of solid sampling GFAAS, which can be used without great problems in glove boxes and hot cells is the slurry technique. In this case two methods have been used. One method uses graphite powder as a diluter, the other is the direct suspension of the sample in a matrix modifier solution. In the case of slurry technique with predilution of the sample with graphite powder, recoveries between 91 and 102% and RSD between 4 and 8% were obtained, whereas in the case of slurry technique with direct suspension of the waste sample recoveries between 91 and 103% and RSD between 14 and 20% for the above mentioned elements were obtained. (orig.)

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Linning, D.L.

1977-01-01

An improvement of the fuel element for a fast nuclear reactor described in patent 15 89 010 is proposed which should avoid possible damage due to swelling of the fuel. While the fuel element according to patent 15 89 010 is made in the form of a tube, here a further metal jacket is inserted in the centre of the fuel rod and the intermediate layer (ceramic uranium compound) is provided on both sides, so that the nuclear fuel is situated in the centre of the annular construction. Ceramic uranium or plutonium compounds (preferably carbide) form the fuel zone in the form of circular pellets, which are surrounded by annular gaps, so that gaseous fission products can escape. (UWI) [de

Nuclear fuel elements and assemblies

International Nuclear Information System (INIS)

Saito, Shozo; Maki, Hideo.

1982-01-01

Purpose: To facilitate the attainment of the uranium enrichment or gadolinia enrichment of a pellet filled in a fuel element. Constitution: The axial length of a pellet filled in a fuel element is set to predetermined sizes according to the uranium enrichment factor, gadolinia enrichment or their combination. Thus, the uranium enrichment factor or gadolinia enrichment can be identified by attaining the axial length of the pellet by using such a pellt. (Kamimura, M.)

MRT fuel element inspection at Dounreay

Energy Technology Data Exchange (ETDEWEB)

Gibson, J.

1997-08-01

To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

Reactor fuel element and fuel assembly

International Nuclear Information System (INIS)

Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

1997-01-01

A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

Effect of fuel particles' size variations on multiplication factor in pebble-bed nuclear reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2005-01-01

The pebble-bed reactor (Pbr) spherical fuel element consists of two radial zones: the inner zone, in which the fissile material in form of the so-called TRISO particles is uniformly dispersed in graphite matrix and the outer zone, a shell of pure graphite. A TRISO particle is composed of a fissile kernel (UO 2 ) and several layers of carbon composites. The effect of TRISO particles' size variations and distance between them on PBR multiplication factor is studied using MCNP code. Fuel element is modelled in approximation of a cubical unit cell with periodic boundary condition. The multiplication factor of the fuel element depends on the size of the TRISO particles due to resonance self-shielding effect and on the inter-particle distance due to inter-kernel shadowing. (author)

French programme for HTR fuel

International Nuclear Information System (INIS)

Gillet, R.M.

1991-01-01

It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

The development of fuel elements for boiling water reactors

International Nuclear Information System (INIS)

Holzer, R.; Kilian, P.

1984-01-01

The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

Graphite-based photovoltaic cells

Science.gov (United States)

Lagally, Max; Liu, Feng

2010-12-28

The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

TRIGA fuel element burnup determination by measurement and calculation

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

2000-01-01

To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

An improved assembly for the transport of fuel elements

International Nuclear Information System (INIS)

Myers, G.

1979-01-01

An improved assembly for the transport and storage of radioactive nuclear fuel elements is described. The fuel element transport canister is of the type in which the fuel elements are submerged in liquid with a self regulating ullage system, so that the fuel elements are always submerged in the liquid even when the assembly is used in one orientation during loading and another orientation during transportation. (UK)

Development of the Fuel Element Database of PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Nurhayati Ramli; Naim Syauqi Hamzah; Nurfazila Husain; Yahya Ismail; Mat Zin Mat Husin; Mohd Fairus Abd Farid

2015-01-01

Since June 28th, 1982, the PUSPATI TRIGA Reactor (RTP) operates safely with an accumulated energy release of about 17,200 MWhr, which corresponds to about 882 g of uranium burn-up. The reactor core has been reconfigured 15th times. Presently, there are 111 TRIGA fuel elements in the core, which 66 of the fuel elements are from the initial criticality while the rest of the fuel elements have been added to compensate the uranium consumption. As 59 % of the fuel elements are older than 30 years old, it is necessary to put the history of every fuel element in a database for easy access of the fuel element movement, inspection results history and integrity status. This paper intends to describe how the fuel element database is developed and related formulae used in determining the RTP fuel element elongation. (author)

Nuclear fuel element

International Nuclear Information System (INIS)

Iwano, Yoshihiko.

1993-01-01

Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

Fuel elements of research reactors in China

International Nuclear Information System (INIS)

Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

1987-01-01

This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

Fuel element

International Nuclear Information System (INIS)

Hirose, Yasuo.

1982-01-01

Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

Fabrication of the Spent Fuel Elements Rack on the ISFSF

International Nuclear Information System (INIS)

Slamet Wiranto; Sigit Purwanto; Safrul, H.

2004-01-01

The Interim Storage For Spent Fuel elements (ISFSF) was designed to be able to store the 33 spent fuel element racks with capacity of 1386 of normal spent fuel elements and 2 racks for 36 of defected ones. Until now, only 9 out of 33 racks of normal spent fuel elements and lout of 2 racks of defected fuel elements are available. Five of them have suffered from corrosion so that they are not fulfilled the requirements of the spent fuel elements storage anymore. Meanwhile, the spent fuel storage racks in the reactor are almost full. It means, the transfer of the spent fuel from reactor spent fuel storage to the ISFSF pool are compulsory needed. Therefore, it is necessary to provide the new ISFSF spent fuel storage rack with better material and fabrication method than the old one. In this design all materials consist of SS 316 L that are welded with the Argon TIG-welding. Right now there has been one new spent fuel storage rack fabricated with capacity of 42 normal spent fuel elements. (author)

Graphite-moderated and heavy water-moderated spectral shift controlled reactors

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

Memory list for the ordering of nuclear fuel elements with UO2 fuel

International Nuclear Information System (INIS)

1977-01-01

The memory list will help to simplify and speed up the technical procedure of fuel element supply for nuclear reactors. Operators of nuclear power plants take great interest in the latest state of thechnology, if sufficiently tested, being applied with regard to material, manufacturing and testing methods. In order to obtain an unlimited availability of the nuclear plant in the future, this application of technology should be taken care of when designing and producing fuel elements. When ordering fuel elements special attention should be drawn to the interdependence of reactor and fuel element with reqard to design and construction, about which, howevers, no further details are given. When ordering fuel elements the operator give the producer all design data of the reactor core and the fuel elements as well as the planned operation mode. He also hands in the respective graphs and the required conditions for design so that a correct and detailed offer can be supplied. An exemplary extent of supply is shown in the given memory list. The regulations required herefore on passing technical material to the fuel element producers have to be established by agreements made by the customer. The order to be given should be itemized as follows: requirements, quality controland quality assurance, warranties and conditions, limits and extent of supply, terms of delivery. (orig./HP) [de

Fuel behavior and fission product release under HTGR accident conditions

International Nuclear Information System (INIS)

Fukuda, K.; Hayashi, K.; Shiba, K.

1990-01-01

In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

Fundamental aspects of nuclear reactor fuel elements

International Nuclear Information System (INIS)

Olander, D.R.

1976-01-01

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO 2 , fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO 2 , radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

Development of electrically heated rods with resistive element of graphite or carbon/carbon composites for simulating transients in nuclear reactors

International Nuclear Information System (INIS)

Polidoro, H.A.

1987-01-01

Thermo-hydraulic problems, in nuclear plants are normally analysed by the use of electrically heated rods. The direct or indirect heater rods are limited in their use because, for high temperatures and high heat flux, the heating element temperature approach its melting point. The use of platinum or tantalum is not economically viable. Graphite and carbon/carbon composites are alternative materials because they are good electrical conductors and have good mechanical properties at high temperatures. Graphite and carbon/carbon composites were used to make heating elements for testing by indirect heating. The swaging process used to reduce the cladding diameter prevented the fabrication of graphite heater rods. Carbon/carbon composite used to make heating elements gave good results up to a heat flux of 100 W/cm 2 . It is easy to verify that this value can be exceeded if the choice of the complementary materials for insulator and cladding improved. (author) [pt

Thermogravimetric and Differential Scanning Calorimetric Behavior of Ball-Milled Nuclear Graphite

Energy Technology Data Exchange (ETDEWEB)

Kim, Eung Seon; Kim, Min Hwan; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi Hyun; Cho, Seung Yon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2013-10-15

An examination was made to characterize the oxidation behavior of ball-milled nuclear graphite powder through a TG-DSC analysis. With the ball milling time, the BET surface area increased with the reduction of particle size, but decreased with the chemisorptions of O{sub 2} on the activated surface. The enhancement of the oxidation after the ball milling is attributed to both increases in the specific surface area and atomic scale defects in the graphite structure. In a high temperature gas-cooled reactor, nuclear graphite has been widely used as fuel elements, moderator or reflector blocks, and core support structures owing to its excellent moderating power, mechanical properties and machinability. For the same reason, it will be used in a helium cooled ceramic reflector test blanket module for the ITER. Each submodule has a seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebbles packed tritium breeder layers, and a reflector layer packed with 1 mm diameter graphite pebbles to reduce the volume of beryllium. The abrasion of graphite structures owing to relative motion or thermal cycle during operation may produce graphite dust. It is expected that graphite dust will be more oxidative than bulk graphite, and thus the oxidation behavior of graphite dust must be examined to analyze the safety of the reactors during an air ingress accident. In this study, the thermal stability of ball-milled graphite powder was investigated using a simultaneous thermogravimeter-differential scanning calorimeter.

Requirements for materials of dispersion fuel elements

International Nuclear Information System (INIS)

Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

1982-01-01

Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

REACTOR FUEL ELEMENTS TESTING CONTAINER

Science.gov (United States)

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Fuel element store

International Nuclear Information System (INIS)

Wieser, R.

1987-01-01

The spherical fuel elements are stored dry in cans. The cans themselves are stacked in parallel storage shafts, which are combined into a rectangular storage space. The storage space is made earthquake-proof by surrounding it with concrete. It consists of a ceiling assembled from several steel parts, which is connected to the floor by support elements. A cooling air ventilation station supplies the individual storage shaft and therefore the cans with cooling air via incoming and outgoing pipes. (DG) [de

A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

Energy Technology Data Exchange (ETDEWEB)

Cuniberti, R; Marullo, G C

1971-02-15

In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

Fundamental aspects of nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Olander, D.R.

1976-01-01

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

Nuclear fuel element

International Nuclear Information System (INIS)

Hirama, H.

1978-01-01

A nuclear fuel element comprises an elongated tube having upper and lower end plugs fixed to both ends thereof and nuclear fuel pellets contained within the tube. The fuel pellets are held against the lower end plug by a spring which is supported by a setting structure. The setting structure is maintained at a proper position at the middle of the tube by a wedge effect caused by spring force exerted by the spring against a set of balls coacting with a tapered member of the setting structure thereby wedging the balls against the inner wall of the tube, and the setting structure is moved free by pushing with a push bar against the spring force so as to release the wedge effect

Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

Energy Technology Data Exchange (ETDEWEB)

Contescu, Cristian I [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burchell, Timothy D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mee, Robert [Univ. of Tennessee, Knoxville, TN (United States)

2015-05-01

This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetime of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.

Structural analysis of reactor fuel elements

International Nuclear Information System (INIS)

Weeks, R.W.

1977-01-01

An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

Further developments of PWR and BWR fuel elements

International Nuclear Information System (INIS)

Sofer, G.A.; Busselman, G.J.; Federico, L.J.

1988-01-01

The performance, safety, and economy of nuclear power plants in inluenced very decisively by the quality of their fuel elements. This is why quality assurance in fuel fabrication has been a factor of great importance from the outset. Operating experince and more stringent performance requirements have resulted in a continuous process of further development of fuel elements, which has been reflected also in lower and lower failure rates and increasingly higher burn-ups. Next to further development also innovation has been an important factor contributing to the present high quality level of fuel elements, which also has allowed fuel cycle costs to be decreased quite considerably. (orig.) [de

Premiering SAFE for Safety Added Fuel Element - 15020

International Nuclear Information System (INIS)

Bhowmik, P.K.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

2015-01-01

The impact of the Fukushima accident has been the willingness to implement passive safety measures in reactor design and to simplify reactor design itself. Within this framework, a new fuel element, named SAFE (Safety Added Fuel Element) based on the concept of accident tolerant fuel, is presented. SAFE is a new type of fuel element cooled internally and externally by light water and with stainless steel as the cladding material. The removal of boron may trigger a series of changes which may simplify the system greatly. A simplified thermal analysis of SAFE shows that the fuel centerline temperature is well below the maximal limit during the normal operation of the plant

Commercial Aspect of Research Reactor Fuel Element Production

International Nuclear Information System (INIS)

Susanto, B.G; Suripto, A

1998-01-01

Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

Recent developments in graphite

International Nuclear Information System (INIS)

Cunningham, J.E.

1983-01-01

Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

Fuel-to-cladding heat transfer coefficient into reactor fuel element

International Nuclear Information System (INIS)

Lassmann, K.

1979-01-01

Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

Nuclear criticality assessment of LEU and HEU fuel element storage

International Nuclear Information System (INIS)

Pond, R.B.; Matos, J.E.

1984-01-01

Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

Attempt to produce silicide fuel elements in Indonesia

International Nuclear Information System (INIS)

Soentono, S.; Suripto, A.

1991-01-01

After the successful experiment to produce U 3 Si 2 powder and U 3 Si 2 -Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x -Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U 3 Si 2 -Al fuel elements, having similar specifications to the ones of U 3 O 8 -Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

Fabrication of fuel elements on the basis of increased concentration fuel composition

International Nuclear Information System (INIS)

Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

2004-01-01

As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

Design and operational behaviour of the SNR-reactor fuel element structure

International Nuclear Information System (INIS)

Dietz, W.; Toebbe, H.

1985-01-01

The fuel element and core concept of a fast breeder reactor is described by the example of the SNR 300 (1st core), and the requirements made on the fuel elements with respect to burnup and neutron dose are listed for existing and projected plants. Irradiation experiments carried out and operational experience gained with fuel elements show that the residence time of the fuel elements is influenced mainly by the stability of shape of the fuel element components. The requirements made with reference to neutron loading for future advanced high-performance fuel elements can not be anticipated from the present state of experience. Besides optimization of fuel element design and checking-out of the limits of operation by PFADFINDERELEMENTE elements, R and D work for the improvement of fuel element materials is also necessary. (orig.) [de

Improvement of graphite crystal analyzer for light elements on X-ray fluorescence holography measurement

Science.gov (United States)

Happo, Naohisa; Hada, Takuma; Kubota, Atsushi; Ebisu, Yoshihiro; Hosokawa, Shinya; Kimura, Koji; Tajiri, Hiroo; Matsushita, Tomohiro; Hayashi, Kouichi

2018-05-01

Using a graphite crystal analyzer, focused monochromatic fluorescent X-rays can be obtained on an X-ray fluorescence holography (XFH) measurement. To measure the holograms of elements lighter than Ti, we improved a cylindrical-type crystal analyzer and constructed a small C-shaped analyzer. Using the constructed C-shaped analyzer, a Ca Kα hologram of a fluorite single crystal was obtained, from which we reconstructed a clear atomic image. The XFH measurements for the K, Ca, and Sc elements become possible using the presently constructed analyzer.

Method 200.12 - Determination of Trace Elements in Marine Waters by StabilizedTemperature Graphite Furnace Atomic Absorption

Science.gov (United States)

This method provides procedures for the determination of total recoverable elements by graphite furnace atomic absorption (GFAA) in marine waters, including estuarine, ocean and brines with salinities of up to 35 ppt.

Nuclear reactor fuel element

International Nuclear Information System (INIS)

D'Eye, R.W.M.; Shennan, J.V.; Ford, L.H.

1977-01-01

Fuel element with particles from ceramic fissionable material (e.g. uranium carbide), each one being coated with pyrolitically deposited carbon and all of them being connected at their points of contact by means of an individual crossbar. The crossbar consists of silicon carbide produced by reaction of silicon metal powder with the carbon under the influence of heat. Previously the silicon metal powder together with the particles was kneaded in a solvent and a binder (e.g. epoxy resin in methyl ethyl ketone plus setting agent) to from a pulp. The reaction temperature lies at 1750 0 C. The reaction itself may take place in a nitrogen atmosphere. There will be produced a fuel element with a high overall thermal conductivity. (DG) [de

The use of graphite for the reduction of void reactivity in CANDU reactors

International Nuclear Information System (INIS)

Min, B.J.; Kim, B.G.; Sim, K-S.

1995-01-01

Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

Domestic nuclear fuels supply: possibility of an independent technology

International Nuclear Information System (INIS)

Cirimello, R.O.

1982-01-01

After considering the different energy sources, their consumption and their respective periods of exploitation, technological considerations in the nuclear fuel field are made. The main subject is the Domestic Supply Project of Embalse Fuel (CANDU type). The different aspects which had to be developed during the realization of this project still under progress, and which are fundamental for the command of the technology, are described: 1) Qualification of the produced fuel elements: fuel elements' characteristics; the reactors' operating parameters, and the prototype fuel elements' characteristics; 2) Development of materials and/or suppliers: the obtainment of UO 2 and its physical properties are considered, as well as those of Zircaloy-4, the development of suppliers and the respective developments for the obtainment of materials such as beryllium, helium and colloidal graphite; 3) Processes development; the following processes are studied and defined: UO 2 pellets fabrication with UO 2 granulated powder; beryllium coating under vaccum; and induction brazing of bearing pads and spacers, end cap and end plate resistance welding and stamping of Zircaloy components, graphite-coating of cladding's internal face; 4) Development of special production equipments; automatic equipment for end cap-to-cladding resistance welding among others. The need for a specific program of quality assurance for nuclear fuels supply is emphasized and the basic criteria are established. The IAEA's quality asssurance requirements are also analyzed. (M.E.L.) [es

Improved nuclear fuel element

International Nuclear Information System (INIS)

Klepfer, H.H.

1974-01-01

A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

Effect of NaX zeolite-modified graphite felts on hexavalent chromium removal in biocathode microbial fuel cells.

Science.gov (United States)

Wu, Xiayuan; Tong, Fei; Yong, Xiaoyu; Zhou, Jun; Zhang, Lixiong; Jia, Honghua; Wei, Ping

2016-05-05

Two kinds of NaX zeolite-modified graphite felts were used as biocathode electrodes in hexavalent chromium (Cr(VI))-reducing microbial fuel cells (MFCs). The one was fabricated through direct modification, and the other one processed by HNO3 pretreatment of graphite felt before modification. The results showed that two NaX zeolite-modified graphite felts are excellent bio-electrode materials for MFCs, and that a large NaX loading mass, obtained by HNO3 pretreatment (the HNO3-NaX electrode), leads to a superior performance. The HNO3-NaX electrode significantly improved the electricity generation and Cr(VI) removal of the MFC. The maximum Cr(VI) removal rate increased to 10.39±0.28 mg/L h, which was 8.2 times higher than that of the unmodified control. The improvement was ascribed to the strong affinity that NaX zeolite particles, present in large number on the graphite felt, have for microorganisms and Cr(VI) ions. Copyright © 2016 Elsevier B.V. All rights reserved.

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

CARA, new concept of advanced fuel element for HWR

International Nuclear Information System (INIS)

Florido, P.C.; Crimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.

1999-01-01

All Argentinean NPPs (2 in operation, 1 under construction), use heavy water as coolant and moderator. With very different reactor concepts (pressure Vessel and CANDU type designs), the fuel elements are completely different in its concepts too. Argentina produces both types of fuel elements at a manufacturing fuel element company, called CONUAR. The very different fuel element's designs produce a very complex economical behavior in this company, due to the low production scale. The competitiveness of the Argentinean electric system (Argentina has a market driven electric system) put another push towards to increase the economical competitiveness of the nuclear fuel cycle. At present, Argentina has a very active Slightly Enriched Uranium (SEU) Program for the pressure vessel HWR type, but without strong changes in the fuel concept itself. Then, the Atomic Energy Commission in Argentina (CNEA) has developed a new concept of fuel element, named CARA, trying to achieve very ambitious goals, and substantially improved the competitiveness of the nuclear option. The ambitious targets for CARA fuel element are compatibility (a single fuel element for all Argentinean's HWR) using a single diameter fuel rod, improve the security margins, increase the burnup and do not exceed the CANDU fabrication costs. In this paper, the CARA concept will be presented, in order to explained how to achieve all together these goals. The design attracted the interest of the nuclear power operator utility (NASA), and the fuel manufacturing company (CONUAR). Then a new Project is right now under planning with the cooperation of three parts (CNEA - NASA - CONUAR) in order to complete the whole development program in the shortest time, finishing in the commercial production of CARA fuel bundle. At the end of the this paper, future CARA development program will be described. (author)

Oxygen reduction kinetics on graphite cathodes in sediment microbial fuel cells.

Science.gov (United States)

Renslow, Ryan; Donovan, Conrad; Shim, Matthew; Babauta, Jerome; Nannapaneni, Srilekha; Schenk, James; Beyenal, Haluk

2011-12-28

Sediment microbial fuel cells (SMFCs) have been used as renewable power sources for sensors in fresh and ocean waters. Organic compounds at the anode drive anodic reactions, while oxygen drives cathodic reactions. An understanding of oxygen reduction kinetics and the factors that determine graphite cathode performance is needed to predict cathodic current and potential losses, and eventually to estimate the power production of SMFCs. Our goals were to (1) experimentally quantify the dependence of oxygen reduction kinetics on temperature, electrode potential, and dissolved oxygen concentration for the graphite cathodes of SMFCs and (2) develop a mechanistic model. To accomplish this, we monitored current on polarized cathodes in river and ocean SMFCs. We found that (1) after oxygen reduction is initiated, the current density is linearly dependent on polarization potential for both SMFC types; (2) current density magnitude increases linearly with temperature in river SMFCs but remains constant with temperature in ocean SMFCs; (3) the standard heterogeneous rate constant controls the current density temperature dependence; (4) river and ocean SMFC graphite cathodes have large potential losses, estimated by the model to be 470 mV and 614 mV, respectively; and (5) the electrochemical potential available at the cathode is the primary factor controlling reduction kinetic rates. The mechanistic model based on thermodynamic and electrochemical principles successfully fit and predicted the data. The data, experimental system, and model can be used in future studies to guide SMFC design and deployment, assess SMFC current production, test cathode material performance, and predict cathode contamination.

Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

International Nuclear Information System (INIS)

Knight, R.W.; Morin, R.A.

1999-01-01

The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U 3 O 8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

Energy Technology Data Exchange (ETDEWEB)

Knight, R.W.; Morin, R.A.

1999-12-01

The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

HTGR fuel element structural design consideration

International Nuclear Information System (INIS)

Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

1987-01-01

The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabilistic stress analysis techniques coupled with probabilistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

HTGR fuel element structural design considerations

International Nuclear Information System (INIS)

Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

1986-09-01

The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

Contribution to the study of can deformations in the fuel elements of gas-graphite reactors during thermal cycling; Contribution a l'etude des deformations des gaines des elements combustibles de reacteur graphite-gaz au cours du cyclage thermique

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M; Boudouresques, B; Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The cans of fuel cartridges used in reactors of the gas-graphite type have either longitudinal fins of variable thickness, short herring-bone fins, or else a mixture of the two. An important test of the strength of these cartridges is their behaviour during thermal cycling carried out in cells reproducing in-pile conditions. It has been observed during with rapid cooling that there occurs a shortening at the base of the fins which can be accompanied in particular by a compression effect at the fin type, which has a tendency to curl, and by a tractive force acting on the body of the can at the ends of the longitudinal fins; this last phenomenon can result in a fracturing of the welds at the extremities or of the ends of the cartridge. This report presents first of all the way in which the stress diagram can be drawn for a can touching the fuel, and then the effect of the ratchet along a fin fixed to a bar with or without grooves. Finally the importance is shown of the test cycling variables (temperature, heating and cooling rates). (authors) [French] Les gaines des cartouches combustibles des reacteurs de la filiere graphite-gaz comportent soit des ailettes longitudinales plus ou moins epaisses, soit de courtes ailettes a chevrons, soit un ensemble des deux. Un test important de la tenue des cartouches, est la tenue au cyclage thermique en cellule pour reproduire le comportement en pile. On a observe au cours des cyclages a refroidissement rapide, un raccourcissement a la base des ailettes qui peut s'accompagner notamment d'une mise en compression du sommet de l'ailette qui a tendance a friser, et d'une traction exercee sur le corps des gaines au bout des ailettes longitudinales; ce dernier phenomene peut se traduire par des ruptures de soudures d'extremites ou des parties terminales de la cartouche. Ce rapport presente d'abord la maniere dont peut etre trace le diagramme des contraintes dans une gaine liee au combustible, puis l'effet du rochet le long d

Statistical estimation of fast-reactor fuel-element lifetime

International Nuclear Information System (INIS)

Proshkin, A.A.; Likhachev, Yu.I.; Tuzov, A.N.; Zabud'ko, L.M.

1980-01-01

On the basis of a statistical analysis, the main parameters having a significant influence on the theoretical determination of fuel-element lifetimes in the operation of power fast reactors in steady power conditions are isolated. These include the creep and swelling of the fuel and shell materials, prolonged-plasticity lag, shell-material corrosion, gap contact conductivity, and the strain diagrams of the shell and fuel materials obtained for irradiated materials at the corresponding strain rates. By means of deeper investigation of these properties of the materials, it is possible to increase significantly the reliability of fuel-element lifetime predictions in designing fast reactors and to optimize the structure of fuel elements more correctly. The results of such calculations must obviously be taken into account in the cost-benefit analysis of projected new reactors and in choosing the optimal fuel burnup. 9 refs

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

Determination of trace elements in paints by direct sampling graphite furnace atomic absorption spectrometry

International Nuclear Information System (INIS)

Bentlin, Fabrina R.S.; Pozebon, Dirce; Mello, Paola A.; Flores, Erico M.M.

2007-01-01

A direct sampling graphite furnace atomic absorption spectrometric (DS-GFAAS) method for the determination of Cd, Pb, Cr, Ni, Co and Cu in paints has been developed. Serigraphy, acrylic and tattoo paints were analysed. Approaches like pyrolysis and atomization temperatures, modifiers and sample mass introduced in the atomizer were studied. Quantification was performed using calibration curves measured with aqueous standard solutions pipetted onto the platform. The sample mass introduced in the graphite tube ranged from 0.02 to 8.0 mg. Palladium was used as modifier for Cd, Pb and Cu, while Mg(NO 3 ) 2 was used for Co. For Ni determination, the graphite platform was covered with carbon powder. The characteristic masses of Cd, Pb, Cr, Ni, Co and Cu were 1.4, 22.5, 7.9, 11.0, 9.6 and 12.5 pg, while the limits of detection were 0.0004, 0.001, 0.03, 0.22, 0.11 and 0.05 μg g -1 of Cd, Pb, Cr, Ni, Co and Cu, respectively. The accuracy was determined by comparison of the results with those obtained by inductively coupled plasma mass spectrometry (ICP-MS) and graphite furnace atomic absorption spectrometry (GFAAS), using liquid sampling of digests. For matrix characterization, major and minor elements (Al, Mg, Ba, Ca, Cr, Cu, Pb, Sr, Ti and Mg) were determined by inductively coupled plasma optical emission spectrometry (ICP OES)

Apparatus and method for assembling fuel elements

International Nuclear Information System (INIS)

Arya, S.P.

1978-01-01

A nuclear fuel element assembling method and apparatus is preferably operable under programmed control unit to receive fuel rods from storage, arrange them into axially aligned stacks of closely monitored length, and transfer the stacks of fuel rods to a loading device for insertion into longitudinal passages in the fuel elements. In order to handle large numbers of one or more classifications of fuel rods or other cylindrical parts, the assembling apparatus includes at least two feed troughs each formed by a pair of screw members with a movable table having a plurality of stacking troughs for alignment with the feed troughs and with a conveyor for delivering the stacks to the loading device, the fuel rods being moved along the stacking troughs upon a fluid cushion. 23 claims, 6 figures

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Upgraded HFIR Fuel Element Welding System

International Nuclear Information System (INIS)

Sease, John D.

2010-01-01

The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

Advancements in the behavioral modeling of fuel elements and related structures

International Nuclear Information System (INIS)

Billone, M.C.; Montgomery, R.O.; Rashid, Y.R.; Head, J.L.

1989-01-01

An important aspect of the design and analysis of nuclear reactors is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system. By understanding the thermomechanical behavior of the different materials which constitute a nuclear fuel element, analysis and predictions can be made regarding the integrity and reliability of fuel element designs. The SMiRT conference series, through the division on fuel elements and the post-conference seminars on fuel element modeling, provided technical forums for the international participation in the exchange of knowledge concerning the thermomechanical modeling of fuel elements. This paper discusses the technical advances in the behavioral modeling of fuel elements presented at the SMiRT conference series since its inception in 1971. Progress in the areas of material properties and constitutive relationships, modeling methodologies, and integral modeling approaches was reviewed and is summarized in light of their impact on the thermomechanical modeling of nuclear fuel elements. 34 refs., 5 tabs

Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B. [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

Postirradiation examination of HTR fuel

International Nuclear Information System (INIS)

Nabielek, H.; Reitsamer, G.; Kania, M.J.

1986-01-01

Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

Energy Technology Data Exchange (ETDEWEB)

Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

2014-12-15

Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Model studying the processes arising during fuel element overheating

International Nuclear Information System (INIS)

Usynin, G.B.; Anoshkin, Yu.I.; Vlasichev, G.N.; Galitskikh, Yu.N.; Semenychev, M.A.

1986-01-01

A calculational technique for studying heating and melting of fuel elements in the BN type reactors during an accident with heat release failure and a simulator with central rod heater intended for out-of-pile experiments is developed. The time rangeof the characteristic melting steps for the most thermally stressed fuel element at the reactor nominal power is calculated. The experimental study of the fuel element melting using a simulator with a tungsten heater has proved that the technique for the simultor and fuel can melting, respectively, is correct. The developed technique is used for determining the geometrical values and operational conditions for experiments with simulators, when quantitative and qualitative characteristics of the process under study are rather close to those natural for fuel elements

Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

Numerical Simulation of Particle Flow Motion in a Two-Dimensional Modular Pebble-Bed Reactor with Discrete Element Method

Directory of Open Access Journals (Sweden)

Guodong Liu

2013-01-01

Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.

Detection and location of leaking TRIGA fuel elements

International Nuclear Information System (INIS)

Bouchey, G.D.; Gage, S.J.

1970-01-01

Several TRIGA facilities have experienced difficulty resulting from cladding failures of aluminum clad TRIGA fuel elements. Recently, at the University of Texas at Austin reactor facility, fission product releases were observed during 250 kW operation and were attributed to a leaking fuel element. A rather extensive testing program has been undertaken to locate the faulty element. The used sniffer device is described, which provides a quick, easily constructed, and extremely sensitive means of locating leaking fuel elements. The difficulty at The University of Texas was compounded by extremely low levels and the sporadic nature of the releases. However, in the more typical situation, in which a faulty element consistently releases relatively large quantities of fission gas, such a device should locate the leak with little difficulty

Biaxial testing for nuclear grade graphite by ball on three balls assessment

International Nuclear Information System (INIS)

Mohd Reusmaazran Yusof; Yusof Abdullah

2012-01-01

Nuclear grade (high-purity) graphite for fuel element and moderator material in Advanced Gas Cooling Reactors (AGR) displays large scatter in strength and a non-linear stress-strain response from the damage accumulation. These responses can be characterized as quasi-brittle behaviour. Current assessments of fracture in core graphite components are based on the linear elastic approximation and thus represent a major assumption. The quasi-brittle behaviour gives challenge to assess the real nuclear graphite component. The selected test method would help to bridge the gap between microscale to macro-scale in real reactor component. The small scale tests presented here can contribute some statistical data to manifests the failure in real component. The evaluation and choice of different solution design of biaxial test will be discussed in this paper. The ball on-three ball test method was used for assessment test follows by numerous of analytical method. The results shown that biaxial strength of the EY9 grade graphite depends on the method used for evaluation. Some of the analytical methods use to calculate biaxial strength were found not to be valid and therefore should not be used to assess the mechanical properties of nuclear graphite. (author)

Quality control in the fuel elements production process

International Nuclear Information System (INIS)

Katanic-Popovic, J.; Spasic, Z.; Djuricis, Lj.

1977-01-01

Recently great attention has been paid at the international level to the analysis of production processes and quality control of fuel and fuel elements with the aim to speed up activity of proposing and accepting standards and measurement methods. IAEA also devoted great interest to these problems appealing to more active participation of all users and producers fuel elements in a general effort to secure successful work of nuclear plants. For adequate and timely participation in future in the establishment and analysis of general requirements and documentation for the control of purchased or self produced fuel elements in out country it is necessary to be well informed and to follow this activity at the international level. (author)

Hot fuel examination facility element spacer wire-wrap machine

International Nuclear Information System (INIS)

Tobias, D.A.; Sherman, E.K.

1989-01-01

Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

Validation of structural design of JHR fuel element

International Nuclear Information System (INIS)

Brisson, S.; Miras, G.; Le Bourdonnec, L.; Lemoine, P.; Anselmet, M.C.; Marelle, V.

2010-01-01

The validation of the structural design of the Jules Horowitz Reactor fuel element was made by the Finite Element Method, starting from the Computer Aided Design. The JHR fuel element is a cylindrical assembly of three sectors composed of eight rolled fuel plates. A roll-swaging process is used to join the fuel plates to three aluminium stiffeners. The hydraulic gap between each plate is 1.95 mm. The JHR fuel assembly is fastened at both ends to the upper and lower endfittings by riveting. The main stresses are essentially thermal loads, imposed on the fuel zone of the plates. These thermal loads result from the nuclear heat flux (W/cm 2 ). The mechanical loads are mainly hydraulic thrust forces. The average coolant velocity is 15 m/s. Seismic effects are also studied. The fuel assembly is entirely modelled by thin shells. The model takes into account asymmetric thermal loads which often appear in Research Reactors. The mechanics of the fuel plates vary in function of the burn up. These mechanical properties are derived from the data sets used in the MAIA code, and the validity of the structure is demonstrable at throughout the life of the fuel. Results concerning displacement are compared to functional criteria, while results concerning stress are compared to RCC-MX criteria. The results of this analysis show that the mechanical and geometrical integrity of the JHR fuel elements is respected for Operating Categories 1 and 2. This paper presents the methodology of this demonstration for the results obtained. (author)

Observation of Compressive Deformation Behavior of Nuclear Graphite by Digital Image Correlation

International Nuclear Information System (INIS)

Kim, Hyunju; Kim, Eungseon; Kim, Minhwan; Kim, Yongwan

2014-01-01

Polycrystalline nuclear graphite has been proposed as a fuel element, moderator and reflector blocks, and core support structures in a very high temperature gas-cooled reactor. During reactor operation, graphite core components and core support structures are subjected to various stresses. It is therefore important to understand the mechanism of deformation and fracture of nuclear graphites, and their significance to structural integrity assessment methods. Digital image correlation (DIC) is a powerful tool to measure the full field displacement distribution on the surface of the specimens. In this study, to gain an understanding of compressive deformation characteristic, the formation of strain field during a compression test was examined using a commercial DIC system. An examination was made to characterize the compressive deformation behavior of nuclear graphite by a digital image correlation. The non-linear load-displacement characteristic prior to the peak load was shown to be mainly dominated by the presence of localized strains, which resulted in a permanent displacement. Young's modulus was properly calculated from the measured strain

Graphite furnace atomic absorption elemental analysis of ecstasy tablets.

Science.gov (United States)

French, Holly E; Went, Michael J; Gibson, Stuart J

2013-09-10

Six metals (copper, magnesium, barium, nickel, chromium and lead) were determined in two separate batches of seized ecstasy tablets by graphite furnace atomic absorption spectroscopy (GFAAS) following digestion with nitric acid and hydrogen peroxide. Large intra-batch variations were found as expected for tablets produced in clandestine laboratories. For example, nickel in batch 1 was present in the range 0.47-13.1 parts per million (ppm) and in batch 2 in the range 0.35-9.06 ppm. Although batch 1 had significantly higher 3,4-methylenedioxy-N-methamphetamine (MDMA) content than batch 2, barium was the only element which discriminated between the two ecstasy seizures (batch 1: 0.19-0.66 ppm, batch 2: 3.77-5.47 ppm). Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

Methodology for substantiation of the fast reactor fuel element serviceability

International Nuclear Information System (INIS)

Tsykanov, V.A.; Maershin, A.A.

1988-01-01

Methodological aspects of fast reactor fuel element serviceability substantiation are presented. The choice of the experimental program and strategies of its realization to solve the problem set in short time, taking into account available experimental means, are substantiated. Factors determining fuel element serviceability depending on parameters and operational conditions are considered. The methodological approach recommending separate studing of the factors, which points to the possibility of data acquisition, required for the development of calculational models and substantiation of fuel element serviceability in pilot and experimental reactors, is described. It is shown that the special-purpose data are more useful for the substantiation of fuel element serviceability and analytical method development than unsubstantial and expensive complex tests of fuel elements and fuel assemblies, which should be conducted only at final stages for the improvement of the structure on the whole

Effect of a Central Graphite Column on a Pebble Flow in a Pebble Bed Core

International Nuclear Information System (INIS)

In, W. K.; Lee, W. J.; Chang, J. H.

2006-01-01

A pebble bed reactor(PBR) uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. The pebble bed core is configured as cylindrical or annular depending on the reactor power. It is well known that an annular core can increase a cores' thermal power. The annular inner core zone is typically filled with movable graphite balls or a fixed graphite column. The first problem with this conventional annular core is that it is difficult to maintain a boundary between the central graphite ball zone and the outer fuel zone. The second problem is that it is expensive to replace the central fixed graphite column after several tens of years of reactor operation. In order to resolve these problems, a PBR with a central graphite column in a low core is invented. This paper presents the effect of the central graphite column on a pebble flow by using the computational fluid dynamics(CFD) code, CFX-10

Gap's dimensions in fuel elements from neutron radiography

International Nuclear Information System (INIS)

Notea, A.; Segal, Y.; Trichter, F.

1985-01-01

Quantitative Nondestructive evaluation (QNDE) is of upmost importance in the design and manufacture of nuclear fuel elements. Accurate non-destructive measurements of gaps, cracks, displacements, etc. supply vital information for optimizing fuel manufacturing. Neutron radiography is a powerful NDT method for examining spent fuel elements. However, it turned out that the extraction of dimensions, especially in the submillimetric range is questionable. In this paper neutron radiography of pellet-to-pellet gaps in fuel elements is modelled and two procedures for dimension extraction are presented. It is shown that for a wide gap the dimension is preferable, extracted from the width of the film profile, while for narrow gaps it is preferable to extract it from the maximum of the density profile

Determination of trace elements in paints by direct sampling graphite furnace atomic absorption spectrometry

Energy Technology Data Exchange (ETDEWEB)

Bentlin, Fabrina R.S. [Instituto de Quimica, Universidade Federal do Rio Grande do Sul, UFRGS, 91501-970 Porto Alegre, RS (Brazil); Pozebon, Dirce [Instituto de Quimica, Universidade Federal do Rio Grande do Sul, UFRGS, 91501-970 Porto Alegre, RS (Brazil)], E-mail: dircepoz@iq.ufrgs.br; Mello, Paola A.; Flores, Erico M.M. [Departamento de Quimica, Universidade Federal de Santa Maria, UFSM, 97105-900 Santa Maria, RS (Brazil)

2007-10-17

A direct sampling graphite furnace atomic absorption spectrometric (DS-GFAAS) method for the determination of Cd, Pb, Cr, Ni, Co and Cu in paints has been developed. Serigraphy, acrylic and tattoo paints were analysed. Approaches like pyrolysis and atomization temperatures, modifiers and sample mass introduced in the atomizer were studied. Quantification was performed using calibration curves measured with aqueous standard solutions pipetted onto the platform. The sample mass introduced in the graphite tube ranged from 0.02 to 8.0 mg. Palladium was used as modifier for Cd, Pb and Cu, while Mg(NO{sub 3}){sub 2} was used for Co. For Ni determination, the graphite platform was covered with carbon powder. The characteristic masses of Cd, Pb, Cr, Ni, Co and Cu were 1.4, 22.5, 7.9, 11.0, 9.6 and 12.5 pg, while the limits of detection were 0.0004, 0.001, 0.03, 0.22, 0.11 and 0.05 {mu}g g{sup -1} of Cd, Pb, Cr, Ni, Co and Cu, respectively. The accuracy was determined by comparison of the results with those obtained by inductively coupled plasma mass spectrometry (ICP-MS) and graphite furnace atomic absorption spectrometry (GFAAS), using liquid sampling of digests. For matrix characterization, major and minor elements (Al, Mg, Ba, Ca, Cr, Cu, Pb, Sr, Ti and Mg) were determined by inductively coupled plasma optical emission spectrometry (ICP OES)

Fabrication of MOX fuel element clusters for irradiation in PWL, CIRUS

International Nuclear Information System (INIS)

Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.

1983-01-01

Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm 2 ; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO 2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

Directory of Open Access Journals (Sweden)

Shigang Lai

2017-01-01

Full Text Available Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancing crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.

A Graphite Oxide Paper Polymer Electrolyte for Direct Methanol Fuel Cells

Directory of Open Access Journals (Sweden)

Ravi Kumar

2011-01-01

Full Text Available A flow directed assembly of graphite oxide solution was used in the formation of free-standing graphene oxide paper of approximate thickness of 100 μm. The GO papers were characterised by XRD and SEM. Electrochemical characterization of the GO paper membrane electrode assembly revealed proton conductivities of 4.1 × 10−2 S cm−1 to 8.2 × 10−2 S cm−1 at temperatures of 25–90°C. A direct methanol fuel cell, at 60°C, gave a peak power density of 8 mW cm−2 at a current density of 35 mA cm−2.

Nuclear fuel element, and method of producing same

International Nuclear Information System (INIS)

Armijo, J.S.; Esch, E.L.

1986-01-01

This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

Energy Technology Data Exchange (ETDEWEB)

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Graphite targets at LAMPF

International Nuclear Information System (INIS)

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

1983-01-01

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

Drying results of K-Basin fuel element 0309M (Run 3)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step

Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

Purification and preparation of graphite oxide from natural graphite

Energy Technology Data Exchange (ETDEWEB)

Panatarani, C., E-mail: c.panatarani@phys.unpad.ac.id; Muthahhari, N.; Joni, I. Made [Instrumentation Systems and Functional Material Processing Laboratory, Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Padjadjaran, Padjadjaran University, Jl. Raya Bandung-Sumedang KM 21, Jatinangor, 45363, Jawa Barat (Indonesia); Rianto, Anton [Grafindo Nusantara Ltd., Belagio Mall Lantai 2, Unit 0 L3-19, Kawasan Mega Kuningan, Kav. B4 No.3, Jakarta Selatan (Indonesia)

2016-03-11

Graphite oxide has attracted much interest as a possible route for preparation of natural graphite in the large-scale production and manipulation of graphene as a material with extraordinary electronic properties. Graphite oxide was prepared by modified Hummers method from purified natural graphite sample from West Kalimantan. We demonstrated that natural graphite is well-purified by acid leaching method. The purified graphite was proceed for intercalating process by modifying Hummers method. The modification is on the reaction time and temperature of the intercalation process. The materials used in the intercalating process are H{sub 2}SO{sub 4} and KMNO{sub 4}. The purified natural graphite is analyzed by carbon content based on Loss on Ignition test. The thermo gravimetricanalysis and the Fouriertransform infrared spectroscopy are performed to investigate the oxidation results of the obtained GO which is indicated by the existence of functional groups. In addition, the X-ray diffraction and energy dispersive X-ray spectroscopy are also applied to characterize respectively for the crystal structure and elemental analysis. The results confirmed that natural graphite samples with 68% carbon content was purified into 97.68 % carbon content. While the intercalation process formed a formation of functional groups in the obtained GO. The results show that the temperature and reaction times have improved the efficiency of the oxidation process. It is concluded that these method could be considered as an important route for large-scale production of graphene.

Irradiation of MEU and LEU test fuel elements in DR 3

International Nuclear Information System (INIS)

Haack, K.

1984-01-01

Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

Energy Technology Data Exchange (ETDEWEB)

Bauer, Alex; Gyenge, Elod L.; Oloman, Colin W. [Department of Chemical and Biological Engineering, The University of British Columbia, 2360 East Mall, Vancouver, BC (Canada)

2007-05-15

Pressed graphite felt (thickness {proportional_to}350 {mu}m) with electrodeposited PtRu (43 g m{sup -2}, 1.4:1 atomic ratio) or PtRuMo (52 g m{sup -2}, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m{sup -2} with PtRuMo at 5500 A m{sup -2} and 353 K while under the same conditions PtRu yielded 1925 W m{sup -2}. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation. (author)

Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

Science.gov (United States)

Bauer, Alex; Gyenge, Előd L.; Oloman, Colin W.

Pressed graphite felt (thickness ∼350 μm) with electrodeposited PtRu (43 g m -2, 1.4:1 atomic ratio) or PtRuMo (52 g m -2, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m -2 with PtRuMo at 5500 A m -2 and 353 K while under the same conditions PtRu yielded 1925 W m -2. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation.

Hydraulic modelling of the CARA Fuel element

International Nuclear Information System (INIS)

Brasnarof, Daniel O.; Juanico, Luis; Giorgi, M.; Ghiselli, Alberto M.; Zampach, Ruben; Fiori, Jose M.; Yedros, Pablo A.

2004-01-01

The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x10 4 and 1,5x10 5 ) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author) [es

Improved nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a metal liner disposed between the cladding and the nuclear fuel material and a high lubricity material in the form of a coating disposed between the liner and the cladding. The liner preferably has a thickness greater than the longest fission product recoil distance and is composed of a low neutron capture cross-section material. The liner is preferably composed of zirconium, an alloy of zirconium, niobium or an alloy of niobium. The liner serves as a preferential reaction site for volatile impurities and fission products and protects the cladding from contact and reaction with such impurities and fission products. The high lubricity material acts as an interface between the liner and the cladding and reduces localized stresses on the cladding due to fuel expansion and cracking of the fuel

HTGR fuel and fuel cycle technology

International Nuclear Information System (INIS)

Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

1977-01-01

Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

Fire and blast safety manual for fuel element manufacture

International Nuclear Information System (INIS)

Ensinger, U.; Koehler, B.; Mester, W.; Riotte, H.G.; Sehrbrock, H.W.

1988-01-01

The manual aims to enable people involved in the planning, operation, supervision, licensing or appraisal of fuel element factories to make a quick and accurate assessment of blast safety. In Part A, technical plant principles are shown, and a summary lists the flammable materials and ignition sources to be found in fuel element factories, together with theoretical details of what happens during a fire or a blast. Part B comprises a list of possible fires and explosions in fuel element factories and ways of preventing them. Typical fire and explosion scenarios are analysed more closely on the basis of experiments. Part B also contains a list and an assessment of actual fires and explosions which have occurred in fuel element factories. Part C contains safety measures to protect against fire and explosion, in-built fire safety, fire safety in plant design, explosion protection and measures to protect people from radiation and other hazards when fighting fires. A distinction is drawn between UO 2 , MOX and HTR fuel elements. (orig./DG) [de

The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

International Nuclear Information System (INIS)

Budisantoso, Edi Trijono; Sardjono, Y.

1996-01-01

The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

In-core fuel element temperature and flow measurment of HFETR

International Nuclear Information System (INIS)

Chen Daolong; Jiang Pei

1988-02-01

The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

Analysis of the ATR fuel element swaging process

International Nuclear Information System (INIS)

Richins, W.D.; Miller, G.K.

1995-12-01

This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

Sodium removal of fuel elements by vacuum distillation

International Nuclear Information System (INIS)

Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

1978-01-01

Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

The Performance of a Direct Borohydride/Peroxide Fuel Cell Using Graphite Felts as Electrodes

Directory of Open Access Journals (Sweden)

Heng-Yi Lee

2017-08-01

Full Text Available A direct borohydride/peroxide fuel cell (DBPFC generates electrical power by recirculating liquid anolyte and catholyte between the stack and reservoirs, which is similar to the operation of flow batteries. To enhance the accessibility of the catalyst layer to the liquid anolyte/catholyte, graphite felts are employed as the porous diffusion layer of a single-cell DBPFC instead of carbon paper/cloth. The effects of the type of anode alkaline solution and operating conditions, including flow rate and temperature of the anolyte/catholyte, on DBPFC performance are investigated and discussed. The durability of the DBPFC is also evaluated by galvanostatic discharge at 0.1 A∙cm−2 for over 50 h. The results of this preliminary study show that a DBPFC with porous graphite electrodes can provide a maximum power density of 0.24 W∙cm−2 at 0.8 V. The performance of the DBPFC drops slightly after 50 h of operation; however, the discharge capacity shows no significant decrease.

Inspection of fuel elements in the cooling pond of a research reactor

International Nuclear Information System (INIS)

Pavlov, S.V.; Mestnikov, A.V.

1992-01-01

Nondestructive testing methods for fuel bundles and fuel elements in the cooling ponds of atomic power plants, using special inspection stands, have come into widespread use during the past decade. This paper describes a methodological stand that was built for the laboratory development of methods and individual units of inspection stands for fuel bundles of RBMK and VVER-1000 reactors. A complex of equipment was developed for the study of irradiated fuel elements, thus creating a methodological base for developing techniques for nondestructive testing of irradiated fuel elements and equipment to obtain information about the state of the fuel elements in a reactor expeditiously. The time required to inspect a fuel element can be shortened using some techniques simultaneously. The length of a fuel element can be measured simultaneously with visual inspection, eddy-current flaw detection can be preformed at the same time as the tranverse size of the fuel element is being determined. 6 refs., 5 figs

HTGR fuel element size reduction system

International Nuclear Information System (INIS)

Strand, J.B.; Cramer, G.T.

1978-06-01

Reprocessing of high-temperature gas-cooled reactor fuel requires development of a fuel element size reduction system. This report describes pilot plant testing of crushing equipment designed for this purpose. The test program, the test results, the compatibility of the components, and the requirements for hot reprocessing are discussed

Modeling of PHWR fuel elements using FUDA code

International Nuclear Information System (INIS)

Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.

2008-01-01

The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated

Space reactor fuel element testing in upgraded TREAT

International Nuclear Information System (INIS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

Space reactor fuel element testing in upgraded TREAT

Science.gov (United States)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

ELOCA: fuel element behaviour during high temperature transients

International Nuclear Information System (INIS)

Sills, H.E.

1979-03-01

The ELOCA computer code was developed to simulate the uniform thermal-mechanical behaviour of a fuel element during high-temperature transients such as a loss-of-coolant accident (LOCA). Primary emphasis is on the diametral expansion of the fuel sheath. The model assumed is a single UO2/zircaloy-clad element with axisymmetric properties. Physical effects considered by the code are fuel expansion, cracking and melting; variation, during the transient, of internal gas pressure; changing fuel/sheath heat transfer; thermal, elastic and plastic sheath deformation (anisotropic); Zr/H 2 O chemical reaction effects; and beryllium-assisted crack penetration of the sheath. (author)

Mechanisms of the initial stage of fuel elements degradation of BN reactor fuel assemblies

International Nuclear Information System (INIS)

Zagorul'ko, Yu.I.; Kashcheev, M.V.; Ganichev, N.S.

2015-01-01

On the base of developed calculational technique numerical evaluation is carried out to the time of fuel element fracture in conditions of loss of sodium flow through fuel element jacket. Data on mechanical properties of steel EhK-164 is used in calculations. Calculations are carried out for different conditions of jacket outer surface cooling: by sodium of 1073 K temperature, by boiling sodium and by sodium in condition of film boiling. It is shown that time to jacket fracture under considered rupture mechanisms essentially depends on fuel element cooling conditions [ru

Process for changing fuel elements of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Fleischmann, R.; Rau, P.

1986-01-01

In order to change fuel elements, a water-filled duct can be installed between the rector pressure vessel and a space for accommodating the fuel elements. The fuel elements are transported there under water by a fuelling machine. The duct is installed as watertight connection closed on all sides between the reactor pressure vessel and a fuel element transport container brought close to it. The fuelling machine works in this duct. (orig./HP) [de

Laser assisted decontamination of nuclear fuel elements

International Nuclear Information System (INIS)

Padma Nilaya, J.; Biswas, Dhruba J.; Kumar, Aniruddha

2010-04-01

Laser assisted removal of loosely bound fuel particulates from the clad surface following the process of pellet loading has decided advantages over conventional methods. It is a dry and noncontact process that generates very little secondary waste and can occur inside a glove box without any manual interference minimizing the possibility of exposure to personnel. The rapid rise of the substrate/ particulate temperature owing to the absorption of energy from the incident laser pulse results in a variety of processes that may lead to the expulsion of the particulates. As a precursor to the cleaning of the fuel elements, initial experiments were carried out on contamination simulated on commonly used clad surfaces to gain a first hand experience on the various laser parameters for which as efficient cleaning can be obtained without altering the properties of the clad surface. The cleaning of a dummy fuel element was subsequently achieved in the laboratory by integrating the laser with a work station that imparted simultaneous rotational and linear motion to the fuel element. (author)

End plug welding of nuclear fuel elements-AFFF experience

International Nuclear Information System (INIS)

Bhatt, R.B.; Singh, S.; Aniruddha Kumar; Amit; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

2004-01-01

Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O 2 fuel elements of various types of nuclear reactors. Fabrication of fuel elements involves pellet fabrication, stack making, stack loading and end plug welding. The requirement of helium bonding gas inside the fuel elements necessitates the top end plug welding to be carried out with helium as the shielding gas. The severity of the service conditions inside a nuclear reactor imposes strict quality control criteria, which demands for almost defect free welds. The top end plug welding being the last process step in fuel element fabrication, any rejection at this stage would lead to loss of effort prior to this step. Moreover, the job becomes all the more difficult with mixed oxide (MOX) as the entire fabrication work has to be carried out in glove box trains. In the case of weld rejection, accepted pellets are salvaged by cutting the clad tube. This is a difficult task and recovery of pellets is low (requiring scrap recovery operation) and also leads to active metallic waste generation. This paper discusses the experience gained at AFFF, in the past 12 years in the area of end plug welding for different types of MOX fuel elements

The advanced neutron source three-element-core fuel grading

International Nuclear Information System (INIS)

Gehin, J.C.

1995-01-01

The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

Spacer grid for fuel elements

International Nuclear Information System (INIS)

Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

1978-01-01

The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 0 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (DG) [de

Spacer grid for fuel elements

International Nuclear Information System (INIS)

Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

1980-01-01

The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (orig.)

RA-3 core with uranium silicide fuel elements

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2000-01-01

Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U 3 O 8 ] fuel elements and type P-06 [from U 3 Si 2 ] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

Design of JMTR high-performance fuel element

International Nuclear Information System (INIS)

Sakurai, Fumio; Shimakawa, Satoshi; Komori, Yoshihiro; Tsuchihashi, Keiichiro; Kaminaga, Fumito

1999-01-01

For test and research reactors, the core conversion to low-enriched uranium fuel is required from the viewpoint of non-proliferation of nuclear weapon material. Improvements of core performance are also required in order to respond to recent advanced utilization needs. In order to meet both requirements, a high-performance fuel element of high uranium density with Cd wires as burnable absorbers was adopted for JMTR core conversion to low-enriched uranium fuel. From the result of examination of an adaptability of a few group constants generated by a conventional transport-theory calculation with an isotropic scattering approximation to a few group diffusion-theory core calculation for design of the JMTR high-performance fuel element, it was clear that the depletion of Cd wires was not able to be predicted accurately using group constants generated by the conventional method. Therefore, a new generation method of a few group constants in consideration of an incident neutron spectrum at Cd wire was developed. As the result, the most suitable high-performance fuel element for JMTR was designed successfully, and that allowed extension of operation duration without refueling to almost twice as long and offer of irradiation field with constant neutron flux. (author)

Fuel elements for pressurised-gas reactors; Elements combustibles des piles a gaz sous pression

Energy Technology Data Exchange (ETDEWEB)

Stohr, J A; Englander, M; Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The design and fabrication of fuel elements for the first CO{sub 2} pressurized reactors have induced to investigate: various cladding materials, natural uranium base fuels, canning processes. The main analogical tests used in connection with the fuel element study are described. These various tests have enabled, among others, the fabrication of the fuel element for the EL2 reactor. Lastly, future solutions for electrical power producing reactors are foreseen. (author)Fren. [French] L'etude et la realisation d'elements combustibles pour les premieres piles a CO{sub 2} sous pression ont conduit a examiner: les divers materiaux de gaine, les combustibles a base d'uranium naturel, les modes de gainage. Les principaux essais analogiques ayant servi au cours de l'etude de la cartouche sont decrits. Ces divers essais ont notamment permis la realisation de la cartouche de la pile EL2. Enfin sont envisagees les solutions futures pour les piles productrices d'energie electrique. (auteur)

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

International Nuclear Information System (INIS)

Anfimov, S.S.

2000-01-01

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

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

Sigma-R: an improved version of the delayed neutron counting device for THTR fuel element verification

International Nuclear Information System (INIS)

Galantucci, R.; Haas, R.; Janssens, A.; Agostini, P.; Becker, L.; Bernede, M.; Caldon, L.; Cordani, G.; Guardini, S.

1989-01-01

SIGMA is a device designed for the Safeguards verification measurements of the 235 U content of THTR (or AVR) pebble fuel elements; it has been in operation for more than 12 years at the HOBEG fuel fabrication plant in HANAU (FRG). The THTR fuel element is a 60 mm diameter sphere composed of particles of a U-Th mixture in a graphite matrix. The device is schematically composed of an irradiation facility where the pebble is irradiated by 252 Cf and a counting facility where the delayed fission neutrons are monitored. The system has recently been thoroughly updated, the modified instrument being called SIGMA-R. SIGMA-R automatically governs the whole charge-discharge, irradiation and counting sequence, collects the measured counts in the microprocessor memory and performs the relevant calculations, including the statistical evaluation. The instructions are given interactively to the inspector through an IRIS terminal. SIGMA-R provides extensive self diagnosis features. The inspector is also informed explicity of malfunctions during routine operation. The instrument is equipped with a permanent memory which facilitates the inspector's work considerably. SIGMA is complementary to the so-called fuel pebble sampling devices (FPSD) which are designed to take a random sample of 10 pebbles out of 1000 directly from the production line. With SIGMA-R it is now possible to measure the 235 U content of THTR pebbles very accurately. The uncertainty essentially is limited to Poisson statistics and allows us to evaluate the 235 U content in routine operation, with a random error down to 0.15% and a systematic uncertainty estimated at 0.25%, the latter coming from calibration and normalization procedures

Drying results of K-Basin fuel element 5744U (Run 4)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Drying Results of K-Basin Fuel Element 2660M (Run 7)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Drying Results of K-Basin Fuel Element 6513U (Run 8)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0

Drying results of K-Basin fuel element 1164M (run 6)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-08-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Assessment of core characteristics during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

International Nuclear Information System (INIS)

Jeong, Chang Joon; Suk, Ho Chun

2002-01-01

A transition from 37-element natural uranium fuel to CANFLEX-NU fuel has been modeled in a 1200-day time-dependent fuel management simulation for a CANDU 6 reactor. The simulation was divided into three parts. The pre-transition period extended from 0 to 300 FPD, in which the reactor was fuelled only with standard 37-element fuel bundles. In the transition period, refueling took place only with the CANFLEX-NU fuel bundle. The transition stage lasted from 300 to 920 FPD, at which point all of the 37-element fuel in the core had been replaced by CANFLEX-NU fuel bundle. In the post-transition phase, refueling continued with CANFLEX-NU fuel until 1200 FPD, to arrive at estimate of the equilibrium core characteristics with CANFLEX-NU fuel. Simulation results show that the CANFLEX-NU fuel bundle has a operational compatibility with the CANDU 6 reactor during the transition core, and also show that the transition core from 37-element natural uranium fuel to CANFLEX-NU can be operated without violating any license limit of the CANDU 6 reactor

Recompressed exfoliated graphite articles

Science.gov (United States)

Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

2013-08-06

This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

Design and main characteristics of HTGR fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Kolesov, V.S.; Permyakov, L.N.; Koshelev, Yu.V.; Mikhajlichenko, L.I.

1983-01-01

Two types of spherical fuel elements and coated particles were investigated under the operating conditions of the high temperature reactors in the Soviet Union (VGR-50 and VG-400). This paper gives the main characteristics of spherical fuel elements (thermal conductivity, static and dynamic strength, wear resistance, release of gaseous fission products, etc.) as determined in test facilities. (author)

Transfer flask for hot active fuel elements

International Nuclear Information System (INIS)

Aubert, Roger; Moutard, Daniel.

1980-01-01

This invention concerns a flask for transporting active fuel elements removed from a nuclear reactor vessel, after only a few days storage and hence cooling, either within a nuclear power station itself or between such a station and a near-by storage area. This containment system is not a flask for conveyance over long and medium distances. Specifically, the invention concerns a transport flask that enables hot fuel elements to be cooled, even in the event of accidents [fr

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

Method to produce carbon-cladded nuclear fuel particles

International Nuclear Information System (INIS)

Sturge, D.W.; Meaden, G.W.

1978-01-01

In the method charges of micro-spherules of fuel element are designed to have two carbon layers, whereby a one aims to achieve a uniform granulation (standard measurement). Two drums are used for this purpose connected behind one another. The micro-spherules coated with the first layer (phenolformaldehyde resin coated graphite particles) leave the first drum and enter the second one. Following the coating with a second layer, the micro-spherules are introduced into a grain size separator. The spherules that are too small are directly recycled into the second drum and those ones that are too large are recycled into the first drum after removing the graphite layers. The method may also be applied to metal cladded particles to manufacture cermet fuels. (RW) [de