Technical ability of new MTR high-density fuel alloys regarding the whole fuel cycle

International Nuclear Information System (INIS)

Durand, J.P.; Maugard, B.; Gay, A.

1998-01-01

The development of new fuel alloys could provide a good opportunity to improve drastically the fuel cycle on the neutronic performances and the reprocessing point of view. Nevertheless, those parameters can only be considered if the fuel manufacture feasibility has been previously demonstrated. As a matter of fact, a MTR work group involving French partners (CEA, CERCA, COGEMA) has been set up in order to evaluate the technical ability of new fuels considering the whole fuel cycle. In this paper CERCA is presenting the preliminary results of UMo and UNbZr fuel plate manufacture, CEA is comparing to U 3 Si 2 the neutronic performances of fuels such as UMo, UN, UNbZr, while COGEMA is dealing with the reprocessing feasibility. (author)

Statistic techniques of process control for MTR type

International Nuclear Information System (INIS)

Oliveira, F.S.; Ferrufino, F.B.J.; Santos, G.R.T.; Lima, R.M.

2002-01-01

This work aims at introducing some improvements on the fabrication of MTR type fuel plates, applying statistic techniques of process control. The work was divided into four single steps and their data were analyzed for: fabrication of U 3 O 8 fuel plates; fabrication of U 3 Si 2 fuel plates; rolling of small lots of fuel plates; applying statistic tools and standard specifications to perform a comparative study of these processes. (author)

Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

International Nuclear Information System (INIS)

Ferrufino, Felipe Bonito Jaldin

2011-01-01

IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

Cost of the external MTR-fuel cycle. (Uranium , reprocessing and related services)

International Nuclear Information System (INIS)

Mueller, H.; Gruber, G.

1991-01-01

This paper points out how the RERTR program has affected NUKEM's fuel supplies for MTRs and how the prices in the External MTR Fuel Cycle have developed during this period. In addition other potential fuel sources and services on the External MTR Fuel Cycle are given. (orig.)

Establishing a LEU MTR fuel manufacturing facility in South Africa

International Nuclear Information System (INIS)

Jamie, R.W.; Kocher, A.

2010-01-01

The South African MTR Fuel Manufacturing Facility was established in the 1970's to supply SAFARI-1 with Fuel Elements and Control Rods. South African capability was developed in parallel with the uranium enrichment program to meet the needs of the Reactor. Further to the July 2005 decision by the South African Governmnent to convert both SAFARI-1 and the Fuel Plant to LEU, the SAFARI-1 phase has been successfully completed and Necsa has commenced with the conversion of the MTR Fuel Manufacturing Facility. In order to establish, validate and qualify the facility, Necsa has entered into a co-operation and technology transfer agreement with AREVA CERCA, the French manufacturer of Research Reactor fuel elements. Past experiences, conversion challenges and the status of the MTR Fuel Facility Project are discussed. On-going co-operation with AREVA CERCA to implement the local manufacture of LEU fuel is explained and elaborated on. (author)

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.)

Transportation of spent MTR fuels

Energy Technology Data Exchange (ETDEWEB)

Raisonnier, D.

1997-08-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

Transportation of spent MTR fuels

International Nuclear Information System (INIS)

Raisonnier, D.

1997-01-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs

The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

International Nuclear Information System (INIS)

Morando, R.A.; Raffaeli, H.A.; Balzaretti, D.E.

1980-01-01

The use of the intermetallic UAl 3 for manufacturing plate type MTR fuel with 20% U 235 enriched uranium and a density of about 20 kg/m 3 is analyzed. The technique used is the dispersion of UAl 3 particles in aluminium powder. The obtainment of the UAl 3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.) [es

MTR loop at the MPR-GA. Siwabessy reactor of Serpong Indonesia for testing of LEU fuel

International Nuclear Information System (INIS)

Arbie, B.; Sunaryadi, D.; Supadi, S.

1991-01-01

The main objective of the MTR-Loop is for testing the specimens of MTR fuel element uprated conditions with respect to the normal conditions of the reactor fuel elements. It is intended to verify the suitability of the fuel elements for operation in a research reactor under preset temperature and pressure conditions. The most important part of the MTR loop is the test section. The fuel elements to be tested are positioned in the test section. For heat removal there is a cooling water flowing through the test section. On this paper the description of the MTR-Loop is described. Installation of the MTR-Loop will be performed in the middle of 1990. In order to facilitate the investigation of fuel behaviour and performance of the new fuel elements the supporting facilities are also already available in the RSG-GAS. (orig.)

A CFD numerical model for the flow distribution in a MTR fuel element

International Nuclear Information System (INIS)

Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho; Angelo, Gabriel

2015-01-01

Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

A CFD numerical model for the flow distribution in a MTR fuel element

Energy Technology Data Exchange (ETDEWEB)

Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho, E-mail: acprado@ipen.br, E-mail: delvonei@ipen.br, E-mail: dpedro_digiovanni_s@hotmail.com, E-mail: fabio@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: jasouza@ipen.br, E-mail: abelchior@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Edvaldo, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil); Angelo, Gabriel, E-mail: gangelo@fei.edu.br [Fundacao Educacional Inaciana (FEI), Sao Bernardo do Campo, SP (Brazil)

2015-07-01

Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

Quality control of nuclear fuel plates using digital image processing techniques

International Nuclear Information System (INIS)

Salinas, Renato; Radd, Ulrich; Coronado, Harold; Olivares, Luis

2003-01-01

The Chilean Atomic Energy Commission (CCHEN) has developed the technology requires to manufacture low enriched uranium-235 nuclear fuel elements used in non-power reactor applications and in research. These fuel plates are assembled in two nuclear facilities located at La Reina (RECH-1) and Lo Aguirre where the present work was developed. Furthermore since high quality standards have been met, these facilities are able to export these nuclear fuel plates to foreign countries. Each MTR fuel elements consists of 16 low enriched uranium silicide (U 3 Si 2 ) fuel plates. A stringent quality assurance program requires among others, homogeneity measurements of uranium surface density values of these fuel plates, which are traditionally accomplished with optical densitometry methods. We have implemented and alternative technique which uses computer vision to determine uranium surface density values in these fuel plates. Both techniques are compared. Advantages of machine vision methods include considerable time saving and a complete quantitative evaluation of uranium densities as compared to the sparse technique involved in the optical densitometry method (Au)

Flow inversion and natural convection in a MTR (Materials Testing Reactor)

International Nuclear Information System (INIS)

Gimenez, M.O.; Clausse, A.

1990-01-01

The thermohydraulic evolution of a refrigerating channel of the MTR (Materials Testing Reactors) RA-6 reactor's core, at the Bariloche Atomic Center, has been studied during the transient caused by the primary system's pump decommissioning. This transient constitutes one of the reactor's operating power boundaries due to the maximum temperature permissible in fuel plates. The problem regarding the thermohydraulic code altered for the rectangular geometry calculation characteristic of the MTR design is analyzed. (Author) [es

A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

Energy Technology Data Exchange (ETDEWEB)

Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

2016-10-21

A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

Performance of the MTR core with MOX fuel using the MCNP4C2 code

International Nuclear Information System (INIS)

Shaaban, Ismail; Albarhoum, Mohamad

2016-01-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

Sipping test on a failed MTR fuel element

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

2002-01-01

This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

International Nuclear Information System (INIS)

Dreesen, K.; Goetze, H.G.; Holst, L.; Gerstenberg, H.; Schreckenbach, K.

2000-01-01

According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was -5 Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

Evaluation of plate type fuel elements by eddy current test method

International Nuclear Information System (INIS)

Frade, Rangel Teixeira

2015-01-01

Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

MTR radiological database for SRS spent nuclear fuel facilities

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

Core conversion from rod to plate type fuel elements in research reactors

International Nuclear Information System (INIS)

Khattab, M.S.; Mina, A.R.

1997-01-01

Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

Corrosion on the fuel plate nucleus based on U3 O8 - Al dispersions

International Nuclear Information System (INIS)

Durazzo, M.

2005-01-01

Samples of MTR type U 3 O 8 - Al dispersion fuel plates meats were corrosion tested in deionized water at different temperatures in the range 30 to 90 deg C. In the tests the cores were exposed to the deionized water by means of an artificially produced cladding defect. The results indicate that the meat corrosion is accompanied by hydrogen evolution. (author)

Flat plate bonded fuel elements: Report number 2, 11 August--10 October 1953

Energy Technology Data Exchange (ETDEWEB)

1953-12-31

Attention has continued to be concentrated on routes employing either wrought uranium or powder metallurgy product for the making of flat plate fuel elements of approximately 0.180-inch uranium metal core thickness bonded to either ribbed or ribless aluminum sheaths. Intermediate goals of the program are to have elements 18 inches long for MTR irradiation tests this fall and to make sufficient advance in the overall program in 1954 so that an initial reactor charge of 15-foot long fuels can be provided as early as possible in 1955. The development of a satisfactory process tube for retaining an assembly of several fuel elements is also required. Uranium of satisfactory quality for fabrication into fuel elements appears to have been produced by the August high alpha rolling at Superior Steel, and it seems likely from the electroplating results that the metal can be employed for electroplating and bonding without such surface preparation as vapor blasting, grinding, or machining. Difficulty in obtaining aluminum components, both sheaths and process tubes, remains a bottleneck in the development program and specifically has delayed work on the wrought metal samples for MTR tests.

Study of the residual porosity in fuel plate cores based on U3O8 - Al dispersions

International Nuclear Information System (INIS)

Durazzo, M.

2005-01-01

The residual porosity in the meat of nuclear dispersion fuel plates, the fabrication voids, explains the corrosion behaviour of the meats when exposed to the water used as coolant and moderator of MTR type research reactors. The fabrication voids also explain variations in irradiation performance of many fuel dispersion for nuclear reactors. To obtain improved corrosion and irradiation performance, we must understand the fabrication factors that control the amount of void volume in fuel plate meats. The purpose of this study was to investigate the void content of aluminum-base dispersion-type U 3 O 8 -Al fuel plates depending on the characteristics of the starting fuel dispersion used to produce the fuel meat, which is fabricated by pressing. The void content depends on the U 3 O 8 concentration. For a particular U 3 O 8 content, the rolling process establishes a constant void concentration, which is called equilibrium porosity. The equilibrium quantity of voids is insensitive to the initial density of the fuel compact. (author)

Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo J.A.; Scuro, Nikolas L.; Andrade, Delvonei A., E-mail: ajcastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum coated fuel plates containing the core of uranium silica (U{sub 3}Si{sub 2}) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. (author)

Development of MTR fuel plate with U-Al dispersion core constituents

International Nuclear Information System (INIS)

Bressiani, Jose Carlos

1979-01-01

This work is a contribution to the development of fuel plates for Research Nuclear Reaction Materials Test Reactors. The plates have the core constituted by dispersions of metallic uranium in aluminum. The main topics of this work are: 1) The preparation of uranium powder with particle sizes in the 53-105μm diameter range; 2) The mixture and cold-pressing of uranium and aluminum powders for different uranium concentrations; 3) The behavior of the dispersions in the roll milling conditions; 4) Blister, radiographic, metallographic and irradiation tests for quality control of the plates. The irradiation test was performed in the IEA-R1 swimming-pool reactor using a prototype with a dispersion of aluminum and natural uranium (45 w/o ), reaching an integrated neutron flux of 8.663 X 10 18 n/cm 2 , no visual changes being noticed after the completion of the experiment. The behavior of the uranium-aluminum reaction for dispersions with 45% w/o uranium also studied. X-ray diffraction experiments showed the formation of UAl 2 UAl 3 and UAl 4 , while energy dispersive analysis of X-rays(EDAX) demonstrated that the diffusion of aluminum in uranium is the mechanism responsible for that reaction. The activation energy for the U-Al reaction was determined by dilatometric experiments yielding 20.2 kcal/mol.The aluminum-uranium reaction reaches an end when extended to 96 h at 600 deg C, namely, when all the uranium is found in the UAl 4 composition. (author)

Rules for the licensing of new experiments in BR2: application to the test irradiation of new MTR-fuels

International Nuclear Information System (INIS)

Joppen, F.

2000-01-01

New types of MTR fuel elements are being developed and require a qualification before routine operation could be authorized. During the test irradiation the new fuel elements .are considered as experimental devices and their irradiation is allowed according to the procedures for experiments. Authorization is based on the advice .of a consultative committee on experiments. This procedure is valid as long as the irradiation is covered by the actual reactor license. An additional license or an amendment is only required if due to the experiment the risk for the workers or the environment is increased in a significant way. A few experimental fuel plates loaded in the primary loop of the reactor will not increase this risk. The source term for potential radioactive releases remains more or less the same. The probability for an accident can be limited by restricting the heat flux and surface temperature. (author)

LEU fuel development at CERCA. Status as of October 1997. Preliminary developments of MTR plates with UMo fuel

International Nuclear Information System (INIS)

Durand, J.P.; Lavastre, Y.; Grasse, M.

1997-01-01

UMo fuels are considered by the RERTR programme because of their higher density as compared to U 3 Si 2 . This paper is focused on the preliminary results about the manufacture feasibility of Uranium/Molybdenum fuel plates carried out by CERCA. A special procedure of casting and heat treatment has been developed in order to get an homogeneous gamma phase of UMo alloy Although U-5%Mo allows to reach densities up to 9.9 U/cm3 with the advanced process developed by CERCA for the high loaded plates, it is not a good candidate on the thermal stability point of view. U-9%Mo alloy seems to gather all the criteria for a good fuel alloy but it is a little less effective on the Uranium density point of view as compared to U-5%Mo alloy. In any case, the preliminary feasibility results are very much encouraging because UMo alloys seem to be compatible with the Aluminium matrix when taking special care while manufacturing. A good compromise could be an intermediate percentage of Molybdenum or the addition of metal traces in order to thermally stabilise 5%Mo. (author)

Status of development and irradiation performance of advanced proliferation resistant MTR fuel at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.; Hassel, H.-W.; Wehner, E.

1985-01-01

This paper describes the current status of development and irradiation performance of fuel elements for Material Test and Research (MTR) Reactors with Medium Enriched Uranium (MEU, ≤ 45 % 235-U) and Low Enriched Uranium (LEU, ≤ 20 % 235-U). (author)

Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

International Nuclear Information System (INIS)

Durazzo, M.

1985-01-01

The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

International Nuclear Information System (INIS)

Sarta Fuentes, Jose A.; Castiblanco, L.A.

2003-01-01

With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)

Conceptual design of next generation MTR

Energy Technology Data Exchange (ETDEWEB)

Nagata, Hiroshi; Yamaura, Takayuki; Naka, Michihiro; Kawamata, Kazuo; Izumo, Hironobu; Hori, Naohiko; Nagao, Yoshiharu; Kusunoki, Tsuyoshi; Kaminaga, Masanori; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Mine, M [Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki (Japan); Yamazaki, S [Kawasaki Heavy Industries, Ltd., Kobe, Hyogo (Japan); Ishikawa, S [NGK Insulators, Ltd., Nagoya, Aichi (Japan); Miura, K [Sukegawa Electric Co., Ltd., Takahagi, Ibaraki (Japan); Nakashima, S [Fuji Electric Co., Ltd., Tokyo (Japan); Yamaguchi, K [Chiyoda Technol Corp., Tokyo (Japan)

2012-03-15

Conceptual design of the high-performance and low-cost next generation materials testing reactor (MTR) which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence

International Nuclear Information System (INIS)

Silva, Clayton Pereira da

2012-01-01

In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U 3 O 8 and U 3 Si 2 later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U 3 Si 2 , meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous chemical treatments (dissolving

Irradiation of full size UMo plates

International Nuclear Information System (INIS)

Vacelet, H.; Lavastre, Y.; Grasse, M.; Sacristan, P.; Languille, A.

1999-01-01

An important development program for a UMo MTR fuel has been launched in France. The goal of the French working group is to develop a high performing and reprocessable fuel before the end of the US return policy. This paper is focussed on the fabrication of full-sized UMo plates with LEU (Low Enriched Enrichment) and their irradiation in OSIRIS reactor which was started on the 22nd of September. The results of the plates inspection are presented here as well as the irradiation conditions. (author)

Comparison Of ²⁵²Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Joshi, Jay Prakash [Los Alamos National Laboratory

2017-03-30

the AmLi source. In this work, two MTR fuel assemblies varying both in size and number of fuel plates were measured using ²⁵²Cf and AmLi active interrogation sources. This paper analyzes time correlated induced fission (TCIF) from fresh MTR fuel assemblies due to ²⁵²Cf and AmLi active interrogation sources.

Analysis of pressure distribution originated over the external plate window of the RA-10 nuclear fuel

International Nuclear Information System (INIS)

Gramajo, M A; Garcia, J.C

2012-01-01

The RA10 is a pool type multipurpose research reactor. The core consists of a rectangular array of MTR fuel type. The refrigeration system at full power and normal operations conditions is carried out by an ascendant flow through the core. To ensure the refrigeration in the sub-channel formed between two adjacent fuels, there is a window orifice over the outer fuel plate. Part of the coolant flow that gets into the fuel will be derived by the window orifice to the sub-channel. Due to the change in the coolant flow direction is necessary to establish the pressure distribution originated over the window In order to achieve this goal a CFD commercial code (FLUENT v6.3.26) was used to perform numerical simulations to obtain the pressure distribution over the window. A quarter of the fuel was modeled using proper symmetry and boundaries conditions (author)

Analysis of fuel management pattern of research reactor core of the MTR type design

International Nuclear Information System (INIS)

Lily Suparlina; Tukiran Surbakti

2014-01-01

Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Sakr, A.M.; Amin, E.H.

2011-01-01

Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

Fuel cell cooler-humidifier plate

Science.gov (United States)

Vitale, Nicholas G.; Jones, Daniel O.

2000-01-01

A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

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

The use of experimental data in an MTR-type nuclear reactor safety analysis

Science.gov (United States)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

The use of experimental data in an MTR-type nuclear reactor safety analysis

International Nuclear Information System (INIS)

Day, S.E.

2006-01-01

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type (i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core. (author)

The use of experimental data in an MTR-type nuclear reactor safety analysis

Energy Technology Data Exchange (ETDEWEB)

Day, S.E

2006-07-01

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type (i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core. (author)

MTR fuel element supply by CERCA through CECCN after the production transfer from NUKEM

International Nuclear Information System (INIS)

Hassel, H.W.

1991-01-01

The transfer of fuel element supply contracts, the corresponding Al-materials, structure parts, documents, uranium metal, customers related know-how, tools and equipment from NUKEM to CERCA has been completed, thus now giving a high flexibility for CERCA's workshop to fabricate and inspect large quantities of several types of fuel elements simultaneously. Based on this fact, on strategic planning for the next couple of years and on the fact that after 10 years of RERTR program the necessary high density fuel has been successfully developed and implemented, 'business as usual' in the field of fabrication has well become possible. The RERTR community should now use the great chance to concentrate all its efforts on problems which still strongly influence the fabrication and the use of MTR fuel elements: supply of enriched uranium,reprocessing capabilities and politics, transports of nuclear materials. (author)

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility

International Nuclear Information System (INIS)

Coragem, Helio Boemer de Oliveira

1980-01-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

International Nuclear Information System (INIS)

Alencar, Donizete Anderson de

2004-01-01

A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

Evaluation of analysis method standardless by WDXRF and EDXRF of aluminum powder used in MTR type fuel

International Nuclear Information System (INIS)

Scapin, Valdirene O.; Salvador, Vera L.R.; Cotrim, Marycel E.B.; Pires, Maria A.F.; Scapin, Marcos A.

2011-01-01

The nuclear fuel used in IEA-R1m reactor at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP) is the MTR type. This fuel is compound of a core (U 3 Si 2 -Al dispersion briquette) wrapped in an aluminum plate with two cladding (superior and inferior) both in aluminum. The fuel element efficiency depends on the quality control of U 3 Si 2 and aluminum. For aluminum should be checked the impurities levels such as Si, Mn, Fe, Co, Cu, Zn and others and Al total . Aiming to provide a quick method, multielemental and non-destructive, the performance of the wavelength dispersive (WDXRF) and energy dispersive (EDXRF) X-ray fluorescence techniques, using the curve instrument sensitivity curve method, also known like standard less analysis, was evaluated. This method allows the determination from the element boron (Z=5) to uranium (Z=92) with concentrations ranging from 0.001 to 99.99% without the need for individual calibration curve and chemical pretreatments in the sample preparation. The results were compared with calibration curve method data, using statistical tests tools. By multivariate analysis of all the experimental data, especially by the discriminant analysis (DA) and cluster analysis (CA), respectively, it was possible to evaluate a correlation between variables of the applied analytical methods could be interpreted in context to qualify the fuels by XRF technique and method standard less. The results showed that the proposed method is satisfactory for both spectrometers; however it was found that the WDXRF presents the greatest conformity degree. (author)

Development of an alternative process for recovery of uranium from rejected plates in the manufacture of MTR type fuel elements

International Nuclear Information System (INIS)

Flores Gonzalez, Jocelyn Natalia

2011-01-01

This work discusses the recovery of enriched uranium in U 235 , from fuel plates rejected during the fuel elements manufacturing process for the La Reina Nuclear Studies Center, RECH-1, CCHEN. The plates have an aluminum based alloy coating, AISI-SAE 6061, with U 3 Si 2 powder distributed evenly inside and dispersed in an aluminum matrix. The high cost of enriched uranium means that it must be recovered from plates rejected in the production process because of non-compliance with the plate specifications, and also because some of them undergo destructive testing, to measure the aluminum coating's thickness on each side of the plate. The thickness of the uranium nucleus is measured as well and the size of the defects on the ends of the plate such as 'dog bone' and 'fish tail', that is, for the purposes of quality control. The first step in the process is carried out by dissolving the aluminum in a hot solution of NaOH in order to release the uranium silicide powder that is insoluble in the soda. A second step involves dissolving the uranium silicide in a hot HNO 3 solution, followed by washing and filtering, and then extracting the SX and analyzing its behavior during this stage. During the process 98.9% of the uranium is recovered together with a solution that is enough for the SX process given the experiences that were carried out in the extraction stage

Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence; Aplicacao da quimiometria para caracterizacao quimica de combustiveis tipo MTR por fluorescencia de raios-X

Energy Technology Data Exchange (ETDEWEB)

Silva, Clayton Pereira da

2012-07-01

In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U{sub 3}Si{sub 2}, meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous chemical

Corrosion of cermet cores of fuel plates for nuclear research reactor

International Nuclear Information System (INIS)

Durazzo, M.; Ramanathan, L.V.

1984-01-01

Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

Nuclear fuel cycle head-end enriched uranium purification and conversion into metal

International Nuclear Information System (INIS)

Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.

1998-01-01

The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)

L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

International Nuclear Information System (INIS)

Huizenga, D. J.

2016-01-01

A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspects of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.

Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

International Nuclear Information System (INIS)

Umbehaun, Pedro Ernesto

2000-01-01

This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

MTR spent fuel back-end - Cogema's long-term commitment

International Nuclear Information System (INIS)

Thomasson, J.

1998-01-01

MTR spent fuel back end has been subject to many reversal and uncertainties in the past 10 years. Until the end of 1988, US obligated materials were subject to the Off site Fuels Policy (OFP). Under this policy, spent fuels were returned to USA, and were reprocessed there. This OFP took end the 31th of December 1988, and Research Reactor's operators had to implement others solutions: On site storage or Reprocessing in Europe. Meanwhile the RERTR Program was leading to a new LEU fuel to replace HEU aluminide. This new silicide fuel has one main drawback: it cannot be reprocessed in working plants without some process main line modifications. Fortunately, a new Research Reactors spent fuels return policy has been set up by the US in the early 1996. This new policy applies to all reactors converted or that have agreed to convert to LEU, and reactors operating with HEU for which no suitable LEU is available. It covers all the spent fuels discharged until 2006/05/12. But after that period of time, each reactor will be fully responsible for its spent fuels. Since the end of 1996, COGEMA is proposing reprocessing services for Aluminides spent fuels, based on the La Hague capability. This COGEMA answer is for the long term, as the La Hague plant has a good load for the coming years, including the first decade of the next century. Further, this activity benefits from a strong R and D support, that allowed fulfilling the evolutive needs of our customers, and gives us the ability to adapt the plant to the future market. Taking advantage of this flexibility, COGEMA offers Research Reactors' operators a long-term commitment. Already two reactors' operators have chosen to contract with COGEMA for the whole life of their reactors. The contracts execution is under progress and the first transportation will take place soon. Beside today's services, COGEMA is involved in R and D activities to support new fuels development enhancing present LEU performances and having the ability to

Dry storage of MTR spent fuel from the Argentine radioisotope production reactor RA-3

International Nuclear Information System (INIS)

Di Marco, A.; Gillaume, E.J.; Ruggirello, G.; Zaweruchi, A.

1996-01-01

The nuclear fuel elements of the RA-3 reactor consist in 19 rectangular fuel plates held in position by two lateral structural plates. The whole assembly is coupled to the lower nozzles that fits in the reactor core grid. The inner plates are 1.5 mm thick, 70.5 mm wide and 655 mm long and the outer plates are 100 mm longer. The fuel plates are formed by a core of an AI-U alloy co-laminated between two plates of Al. Enrichment is 90% 235 U. After being extracted from the reactor, the fuel elements have been let to cool down in the reactor storage pool and finally moved to the storage facility. This facility is a grid of vertical underground channels connected by a piping system. The system is filled with processed and controlled water. At the present the storage capacity of the facility is near to be depleted and some indications of deterioration of the fuel elements has been detected. Due to the present status of the facility and the spent fuel stored there, a decision has been taken to proceed to modify the present underwater storage to dry storage. The project consist in: a) Decontamination and conditioning of the storage channels to prepare them for dry storage. b) Disassembly of the fuel elements in hot cells in order to can only the active fuel plates in an adequate tight canister. c) The remnant structural pieces will be treated as low level waste. (author). 10 figs

Modeling RERTR experimental fuel plates using the PLATE code

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Snelgrove, J.L.; Brazener, R.A.

2003-01-01

Modeling results using the PLATE dispersion fuel performance code are presented for the U-Mo/Al experimental fuel plates from the RERTR-1, -2, -3 and -5 irradiation tests. Agreement of the calculations with experimental data obtained in post-irradiation examinations of these fuels, where available, is shown to be good. Use of the code to perform a series of parametric evaluations highlights the sensitivity of U-Mo dispersion fuel performance to fabrication variables, especially fuel particle shape and size distributions. (author)

Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence; Aplicacao da quimiometria para caracterizacao quimica de combustiveis tipo MTR por fluorescencia de raios-X

Energy Technology Data Exchange (ETDEWEB)

Silva, Clayton Pereira da

2012-07-01

In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U{sub 3}Si{sub 2}, meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous

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)

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

International Nuclear Information System (INIS)

Schneider, B.; Hofmann, G.

1994-09-01

Most of the spent fuel assemblies are temporarily stored in the flooded fuel ponds AB 1 and AB 2 of the RFR, and some are still in the reactor core. The conditioning task described here is part of the RFR spent fuel management concept and covers the safe emplacement of the spent fuel elements in the CASTOR MTR 2 shipping containers and the sealing of the containers in compliance with the nuclear licence issued for the conditioning task. The transfer of the spent fuel assemblies from the present wet storage conditions to the dry storage conditions in the CASTOR MTR 2 containers is done by a mobile manipulation equipment consisting essentially of the transfer sluice gate and a transfer container. Subsequent to conditioning, the shipping containers are to be transported to a licensed intermediate storage facility to await their transport to a national radwaste repository. The technical handling tools for the transfer and manipulation are briefly described, as well as the process steps involved, putting emphasis on the detailed description of processes and the accompanying time frame, so that the conditioning task can be incorporated into the work plan of the entire project. The report further presents the EDP concept established for the task, including the required data archivation and documentation. (orig.) [de

Preparation of U3O8 powder for MTR type fuel from ammonium uranyl carbonate

International Nuclear Information System (INIS)

Marcondes, G.H.; Riella, H.G.

1990-08-01

In this paper it is described the research done at IPEN-CNEN/SP on the preparation of U 3 O 8 powder from calcination of the AUC, with appropriate characteristics to be used as dispersoid for MTR type fuel. The calcination in air of the AUC leads a U 3 O 8 powder that is further processed to obtain a powder with density and particle size as especifications. The important process parameters are here discussed with the variation AUC calcination temperature and sintering time of the U 3 O 8 powder. (author) [pt

Parameters calculation of fuel assembly with complex geometry

International Nuclear Information System (INIS)

Wu Hongchun; Ju Haitao; Yao Dong

2006-01-01

The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)

MTR core loading pattern optimization using burnup dependent group constants

Directory of Open Access Journals (Sweden)

Iqbal Masood

2008-01-01

Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

Neutronic analysis of the conversion of HEU to LEU fuel for a 5-MW MTR core

International Nuclear Information System (INIS)

Pazirandeh, A.; Bartsch, G.

1987-01-01

In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% 235 U) is presented. In order to assess the performance of the core with the LEU ( 235 U loadings were examined. The core consists of 22 standard fuel elements (SFEs) and 6 control fuel elements (CFEs). Each fuel elements has 18 curved plates of which two end plates are dummies. Initial 235 U content is 195 g 235 U/SFE and 9.7 g 235 U/CFE or /PFE. In all calculations the permitted changes to the fuel elements are (a) 18 active plates per SFE, (b) fuel plates assumed to be flat, and (c) 8 or 9 active plates per CFE

Status of high-density fuel plate fabrication

International Nuclear Information System (INIS)

Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

1991-01-01

Progress has continued on the fabrication of fuel plates with equivalent fuel zone loadings approaching 9 gU/cm 3 . Through hot isostatic pressing (HIP), successful diffusion bonds have been made with 1100 Al and 6061 Al alloys. Although additional study is necessary to optimize the procedure, these bonds demonstrated the most critical processing step for proof-of-concept hardware. Two types of prototype highly loaded fuel plates have been fabricated. The first is a fuel plate in which 0.030-in. (0.76-mm) uranium compound wires are bonded within an aluminum cladding; the second, a dispersion fuel plate with uniform cladding and fuel zone thickness. The successful fabrication of these fuel plates derives from the unique ability of the HIP process to produce diffusion bonds with minimal deformation. (orig.)

Early detection of coolant boiling in research reactors with MTR-type fuel

International Nuclear Information System (INIS)

Kozma, R.; Turkcan, E.; Verhoef, J.P.

1992-10-01

A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

Reactivity worth of the thermal column of a MTR type swimming pool research reactor using low enriched uranium fuel

International Nuclear Information System (INIS)

Ali Khan, L.; Ahmad, N.

2002-01-01

The reactivity worth of the thermal column of a typical MTR type swimming pool research reactor using low enriched uranium fuel has been determined by modeling the core using standard computer codes. It was also measured experimentally by operating the reactor in the stall and open ends. The calculated value of the reactivity worth of the thermal column is about 14% greater than the experimentally determined value

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

Energy Technology Data Exchange (ETDEWEB)

Drera, Saleem S., E-mail: saleem.drera@gmail.com [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); Hofman, Gerard L. [Argonne National Laboratory, Chicago, IL 60439 (United States); Kee, Robert J. [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); King, Jeffrey C. [Metallurgical and Materials Engineering, Colorado School of Mines, Golden, CO 80401 (United States)

2014-10-15

Highlights: • This article presents a cellular automata (CA) algorithm to synthesize the growth of intermetallic interaction layers in U–Mo/Al dispersion fuel. • The method utilizes a 3D representation of the fuel, which is discretized into separate voxels that can change identy based on derived CA rules. • The CA model is compared to ILT measurements for RERTR experimental data. • The primary objective of the model is to synthesize three-dimensional microstructures that can be used in subsequent thermal and mechanical modeling. • The CA model can be used for predictive analysis. For example, it can be used to study the dependence of temperature on interaction layer growth. - Abstract: Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium–molybdenum (U–Mo) particles within an aluminum matrix. Fresh U–Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction–diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

Effect of core configuration on the burnup calculations of MTR research reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Amin, E.H.; Sakr, A.M.

2014-01-01

Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

Energy Technology Data Exchange (ETDEWEB)

Barrett, T.R.; Harrison, R. [UKAEA, Nuclear Materials Control Dep., Dounreay (United Kingdom)

1997-07-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

International Nuclear Information System (INIS)

Barrett, T.R.; Harrison, R.

1997-01-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

Prevention of criticality accidents. Fuel elements storage

International Nuclear Information System (INIS)

Canavese, S.I.; Capadona, N.M.

1990-01-01

Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author) [es

Low-enriched uranium-molybdenum fuel plate development

International Nuclear Information System (INIS)

Wiencek, T.C.; Prokofiev, I.G.

2000-01-01

To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)

Evolution of fuel plate parameters during deformation in rolling

Energy Technology Data Exchange (ETDEWEB)

Durazzo, M., E-mail: mdurazzo@ipen.br [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Vieira, E.; Urano de Carvalho, E.F. [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Riella, H.G. [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Chemical Engineering Department, Santa Catarina Federal University, Florianópolis (Brazil)

2017-07-15

The Nuclear and Energy Research Institute – IPEN/CNEN-SP routinely produces the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U{sub 3}Si{sub 2}-Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed based on information obtained from literature, which was used as a premise for defining the current manufacturing procedures, according to a methodology with an essentially empirical character. Despite the current rolling process being perfectly stable and highly reproducible, it is not well characterized and is therefore not fully known. The objective of this work is to characterize the rolling process for producing dispersion fuel plates. Results regarding the evolution of the main parameters of technological interest, after each rolling pass, are presented. Some defects that originated along the fuel plate deformation during the rolling process were characterized and discussed. The fabrication procedures for manufacturing the fuel plates are also presented. - Highlights: •Evolution of defects when manufacturing dispersion fuel plates. •Aspects of dispersion fuel plates fabrication. •What happen during the manufacturing of dispersion fuel plates? •Clarifying the deformation of fuel plates by rolling.

Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection; Avaliacao de integridade de revestimentos de combustiveis de reatores de pesquisa e teste de materiais utilizando o ensaio de correntes parasitas

Energy Technology Data Exchange (ETDEWEB)

Alencar, Donizete Anderson de

2004-07-01

A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

International Nuclear Information System (INIS)

Weiss, M.

2001-01-01

As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

Planning a new research reactor for AECL: The MAPLE-MTR concept

International Nuclear Information System (INIS)

Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

1992-01-01

AECL Research is assessing its needs and options for future irradiation research facilities. A planning team has been assembled to identify the irradiation requirements for AECL's research programs and compile options for satisfying the irradiation requirements. The planning team is formulating a set of criteria to evaluate the options and will recommend a plan for developing an appropriate research facility. Developing the MAPLE Materials Test Reactor (MAPLE-MTR) concept to satisfy AECL's irradiation requirements is one option under consideration by the planning team. AECL is undertaking this planning phase because the NRU reactor is 35 years old and many components are nearing the end of their design life. This reactor has been a versatile facility for proof testing CANDU components and fuel designs because the CANDU irradiation environment was simulated quite well. However, the CANDU design has matured and the irradiation requirements have changed. Future research programs will emphasize testing CANDU components near or beyond their design limits. To provide these irradiation conditions, the NRU reactor needs to be upgraded. Upgrading and refurbishing the NRU reactor is being considered, but the potentially large costs and regulatory uncertainties make this option very challenging. AECL is also developing the MAPLE-MTR concept as a potential replacement for the NRU reactor. The MAPLE-MTR concept starts from the recent MAPLE-X10 design and licensing experience and adapts this technology to satisfy the primary irradiation requirements of AECL's research programs. This approach should enable AECL to minimize the need for major advances in nuclear technology (e.g., fuel design, heat transfer). The preliminary considerations for developing the MAPLE-MTR concept are presented in this report. A summary of AECL's research programs is presented along with their irradiation requirements. This is followed by a description of safety criteria that need to be taken into

Description of ECRI (CNEA'S MTR fuel fabrication plant)

International Nuclear Information System (INIS)

Echenique, P.; Fabro, J.; Podesta, D.; Restelli, M.; Rossi, G.; Alvarez, L.; Adelfang, P.

2002-01-01

The ECRI Plant is dedicated to the development and fabrication of high-density fuel elements and targets for 99 Mo. In this sector had been done the start up Fuel Elements for the Reactors of Peru, Iran, Algeria and Egypt. All of them were made with U 3 O 8 . The targets for 99 Mo using HEU were fabricated too in the last years. The new material of high-density for Fuel Elements as U 3 Si 2 were done in this sector, three prototypes were fabricated, two are still under irradiation. (P06 and P07). As new developments we are working with U-Mo (7%) Fuel Plates with both material Korean and HMD. This work is under the RERTR Program and two fuel elements, manufactured by us, with both powders, will be irradiated in Petten. For 99 Mo targets, we are fabricating miniplates of LEU with an AlUx powder by pulvi-metallurgy technique. And it is under development the foils targets under the RERTR Program. A general view of the fabrication facilities and control sector will be shown. The different operations that are done in each sector will be explained. All our activities will be certified under the ISO 9000 and we are working hard to get it in the middle of 2003. (author)

Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

International Nuclear Information System (INIS)

Blumenfeld, P.E.

1995-08-01

Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

General description and production lines of the Egyptian fuel manufacturing pilot plant

International Nuclear Information System (INIS)

Zidan, W.I.; Elseaidy, I.M.

1999-01-01

The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a new facility, producing an MTR-type fuel elements required for the Egyptian Second Research Reactor, ETRR-2, as well as other plates or elements for an external clients with the same type and enrichment percent or lower, (LEU). General description is presented. The production lines in FMPP, which begin from uranium hexaflouride (UF 6 , 19.7±0.2 % U 235 by wt), aluminum powder, and nuclear grade 6061 aluminium alloy in sheets, bars, and rods with the different heat treatments and dimensions as a raw materials, are processed through a series of the manufacturing, inspection, and quality control plan to produce the final specified MTR-type fuel elements. All these processes and the product control in each step are presented. The specifications of the final product are presented. (author)

Flow-induced plastic collapse of stacked fuel plates

Energy Technology Data Exchange (ETDEWEB)

Davis, D C; Scarton, H A

1985-03-01

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental reactors such as the ORNL High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. A methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.

Finite element analysis of advanced neutron source fuel plates

International Nuclear Information System (INIS)

Luttrell, C.R.

1995-08-01

The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles

HEU and LEU MTR fuel elements as target materials for the production of fission molybdenum

International Nuclear Information System (INIS)

Sameh, A.A.; Bertram-Berg, A.

1993-01-01

The processing of irradiated MTR-fuels for the production of fission nuclides for nuclear medicine presents a significantly increasing task in the field of chemical separation technology of high activity levels. By far the most required product is MO-99, the mother nuclide of Tc-99m which is used in over 90% of the organ function tests in nuclear medicine. Because of the short half life of Mo-99 (66 h) the separation has to be carried out from shortly cooled neutron irradiated U-targets. The needed product purity, the extremely high radiation level, the presence of fission gases like xenon-133 and of volatile toxic isotopes such as iodine-131 and its compounds in kCi-scale require a sophisticated process technology

Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

International Nuclear Information System (INIS)

Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

2010-01-01

As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

The French UMo group contribution to new LEU fuel development

International Nuclear Information System (INIS)

Hamy, J.M.; Lemoine, P.; Huet, F.; Jarousse, C.; Emin, J.L.

2005-01-01

The French UMo Group was based on a close collaboration between CEA and AREVA's companies strongly involved in the MTR field. The aim of this program was to deliver industrially a high performance LEU UMo fuel able to be reprocessed, and suitable for a wide range of Research Reactor, covering the expected needs for MTR next generation. Since 1999, the program has been focused on industrial aspects with the intention to deal with the whole fuel cycle: manufacturing, irradiation behaviour, fuel characterisation, code development and reprocessing validation. It has been based on the fabrication of full-sized U-7%Mo fuel plates with a density up to 8 gU/cm 3 . The dedicated and advanced R and D means provided by the CEA have been used intensively with the contribution of HFR and BR2 facilities in Europe. This paper presents a synthesis of the program and the corresponding significant results obtained. These results have played a major role as regards the UMo dispersion fuel qualification route by issuing, for the first time, evidence of severe performance limitations. Consequently, the global international effort to develop and qualify a high density LEU UMo fuel has been definitively re-routed and forced to overcome these discrepancies by exploring new technical solutions. A French extended program sustained by a CEA and CERCA collaboration has been launched in 2004 in order to develop a suitable UMo fuel solution. UMo dispersion and monolithic fuel are both investigated through three new full-sized plate irradiations planned in OSIRIS. (author)

Modeling of high-density U-MO dispersion fuel plate performance

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Rest, J.; Snelgrove, J.L.

2002-01-01

Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

Fuel cell end plate structure

Science.gov (United States)

Guthrie, Robin J.; Katz, Murray; Schroll, Craig R.

1991-04-23

The end plates (16) of a fuel cell stack (12) are formed of a thin membrane. Pressure plates (20) exert compressive load through insulation layers (22, 26) to the membrane. Electrical contact between the end plates (16) and electrodes (50, 58) is maintained without deleterious making and breaking of electrical contacts during thermal transients. The thin end plate (16) under compressive load will not distort with a temperature difference across its thickness. Pressure plate (20) experiences a low thermal transient because it is insulated from the cell. The impact on the end plate of any slight deflection created in the pressure plate by temperature difference is minimized by the resilient pressure pad, in the form of insulation, therebetween.

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Energy Technology Data Exchange (ETDEWEB)

Wachs, D.M., E-mail: daniel.wachs@inl.gov [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Robinson, A.B.; Rice, F.J. [Idaho National Laboratory, Characterization and Advanced PIE Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Kraft, N.C.; Taylor, S.C. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lillo, M. [Idaho National Laboratory, Nuclear Systems Design and Analysis Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Woolstenhulme, N.; Roth, G.A. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm{sup 3} peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Bipolar plates for PEM fuel cells

Science.gov (United States)

Middelman, E.; Kout, W.; Vogelaar, B.; Lenssen, J.; de Waal, E.

The bipolar plates are in weight and volume the major part of the PEM fuel cell stack, and are also a significant contributor to the stack costs. The bipolar plate is therefore a key component if power density has to increase and costs must come down. Three cell plate technologies are expected to reach targeted cost price levels, all having specific advantages and drawbacks. NedStack has developed a conductive composite materials and a production process for fuel cell plates (bipolar and mono-polar). The material has a high electric and thermal conductivity, and can be processed into bipolar plates by a proprietary molding process. Process cycle time has been reduced to less than 10 s, making the material and process suitable for economical mass production. Other development work to increase material efficiency resulted in thin bipolar plates with integrated cooling channels, and integrated seals, and in two-component bipolar plates. Total thickness of the bipolar plates is now less than 3 mm, and will be reduced to 2 mm in the near future. With these thin integrated plates it is possible to increase power density up to 2 kW/l and 2 kW/kg, while at the same time reducing cost by integrating other functions and less material use.

Drying studies of simulated DOE aluminum plate fuels

International Nuclear Information System (INIS)

Lords, R.E.; Windes, W.E.; Crepeau, J.C.; Sidwell, R.W.

1996-01-01

Experiments have been conducted to validate the Idaho National Engineering Laboratory (INEL) drying procedures for preparation of corroded aluminum plate fuel for dry storage in an existing vented (and filtered) fuel storage facility. A mixture of hydrated aluminum oxide bound with a clay was used to model the aluminum corrosion product and sediment expected in these Department of Energy (DOE) owned fuel types. Previous studies demonstrated that the current drying procedures are adequate for removal of free water inside the storage canister and for transfer of this fuel to a vented dry storage facility. However, using these same drying procedures, the simulated corrosion product was found to be difficult to dry completely from between the aluminum clad plates of the fuel. Another related set of experiments was designed to ensure that the fuel would not be damaged during the drying process. Aluminum plate fuels are susceptible to pitting damage on the cladding that can result in a portion of UAl x fuel meat being disgorged. This would leave a water-filled void beneath the pit in the cladding. The question was whether bursting would occur when water in the void flashes to steam, causing separation of the cladding from the fuel, and/or possible rupture. Aluminum coupons were fabricated to model damaged fuel plates. These coupons do not rupture or sustain any visible damage during credible drying scenarios

Characterization and testing of monolithic RERTR fuel plates

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D.; Jue, J.F.; Burkes, D.E. [Idaho National Lab., Idaho Falls, ID (United States)

2007-07-01

Monolithic fuel plates are being developed as a LEU (low enrichment uranium) fuel for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIP-made samples indicates that the foil/cladding interaction behavior depends on the Mo content in the UMo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process. (authors)

JHR. A high performance MTR under construction for a sustainable nuclear energy

International Nuclear Information System (INIS)

Iracane, Daniel; Cordier, Pierre-Yves

2009-01-01

The Access to an up-to-date Material Testing Reactor (MTR) is essential to support a sustainable nuclear energy, meeting industry and public needs, and keeping a high level of scientific expertise. This includes services to existing and coming reactor technologies for major stakes such as safety and competitiveness, lifetime management, operation optimization, development of innovative structural material and fuel required for future systems (innovative Gen III, Gen IV, fusion...), etc. The JHR copes with this context. Design phase has been completed by the end of 2005 and JHR is now under construction. Start of operation is scheduled in 2014. As a new MTR taking benefit of a large available worldwide experience, JHR offers new major experimental capability that will be presented. JHR will be operated within an international users' consortium that will guarantee effective and cost-effective operation. This innovative way to operate a MTR, as a user-facility for the benefit of industry and public bodies, will be presented. (author)

Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

International Nuclear Information System (INIS)

Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

2010-01-01

Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

Dual fuel gradients in uranium silicide plates

Energy Technology Data Exchange (ETDEWEB)

Pace, B.W. [Babock and Wilcox, Lynchburg, VA (United States)

1997-08-01

Babcock & Wilcox has been able to achieve dual gradient plates with good repeatability in small lots of U{sub 3}Si{sub 2} plates. Improvements in homogeneity and other processing parameters and techniques have allowed the development of contoured fuel within the cladding. The most difficult obstacles to overcome have been the ability to evaluate the bidirectional fuel loadings in comparison to the perfect loading model and the different methods of instilling the gradients in the early compact stage. The overriding conclusion is that to control the contour of the fuel, a known relationship between the compact, the frames and final core gradient must exist. Therefore, further development in the creation and control of dual gradients in fuel plates will involve arriving at a plausible gradient requirement and building the correct model between the compact configuration and the final contoured loading requirements.

Flow velocity calculation to avoid instability in a typical research reactor core

International Nuclear Information System (INIS)

Oliveira, Carlos Alberto de; Mattar Neto, Miguel

2011-01-01

Flow velocity through a research reactor core composed by MTR-type fuel elements is investigated. Core cooling capacity must be available at the same time that fuel-plate collapse must be avoided. Fuel plates do not rupture during plate collapse, but their lateral deflections can close flow channels and lead to plate over-heating. The critical flow velocity is a speed at which the plates collapse by static instability type failure. In this paper, critical velocity and coolant velocity are evaluated for a typical MTR-type flat plate fuel element. Miller's method is used for prediction of critical velocity. The coolant velocity is limited to 2/3 of the critical velocity, that is a currently used criterion. Fuel plate characteristics are based on the open pool Australian light water reactor. (author)

OSIRIS, a MTR adapted and well fitted to LEU utilization qualification and development

International Nuclear Information System (INIS)

Barnier, M.; Beylot, J.P.

1984-01-01

The MTR OSIRIS has been successfully operated for 4 years using the ''Caramel'' low enriched uranium dioxyde fuel for the whole core loading. In the first part we examine the performance and operating experience obtained up to the present time with ''Caramel''. In a second part the paper discusses the results of the calculations for a complete OSIRIS core loaded with 20 % silicide fuel and makes a comparison with UAl 93 % and ''Caramel'' 7 % fuels. (author)

Fuel performance analysis for the HAMP-1 mini plate test

Energy Technology Data Exchange (ETDEWEB)

Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

Creep analysis of fuel plates for the Advanced Neutron Source

International Nuclear Information System (INIS)

Swinson, W.F.; Yahr, G.T.

1994-11-01

The reactor for the planned Advanced Neutron Source will use closely spaced arrays of fuel plates. The plates are thin and will have a core containing enriched uranium silicide fuel clad in aluminum. The heat load caused by the nuclear reactions within the fuel plates will be removed by flowing high-velocity heavy water through narrow channels between the plates. However, the plates will still be at elevated temperatures while in service, and the potential for excessive plate deformation because of creep must be considered. An analysis to include creep for deformation and stresses because of temperature over a given time span has been performed and is reported herein

Development of hold down plate of INGLE fuel assembly

International Nuclear Information System (INIS)

Kim, Hyeong Koo; Kim, Kyu Tae

1996-07-01

Hold down plate for the INGLE fuel which has been designed for high performance in the standpoints of thermal margin and structural integrity compared to current fuel for YGN 3/4 and UCN 3/4 has been developed and its structural integrity has been verified based on the eh stress analysis. The design feature of the developed hold down plate has not only perfect compatibility with the reactor internals of Korea standard reactor, but also brand-new locking mechanism between upper tie plate and guide tubes. This locking mechanism introduced to the INGLE fuel provides very simple and reliable reconstitutability. In this report, finite element stress analysis with the aid of the ANSYS code as a solver and the MSC/PATRAN code as a pre and post processor were performed to verify structural integrity of the hold down plate considering various load cases which seem to be applied to the hold down plate during its lifetime. Based on the analysis results, the developed hold down plate for INGLE fuel sustains structural integrity under considered load conditions. 3 tabs., 16 figs., 9 refs. (Author)

Design of the Flow Plates for a Dual Cooled Fuel Assembly

International Nuclear Information System (INIS)

Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

2009-01-01

In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plate provides flow holes to direct the heated coolant into/out of the fuel assembly and structural intensity to insure that the fuel rod is axially restrained within the spacer grids. So, flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape. Because the flow holes' area of a flow plate affects pressure drop, the flow holes' area must be larger than/equal to that of conventional flow plates. And design criterion of the TEP/BEP says that the flow plate should withstand a 22.241 kN axial load during handling lest a calculated stress intensity should exceed the Condition I allowable stress. In this paper, newly designed flow plates of a TEP/BEP are suggested and stress analysis is conducted to evaluate strength robustness of the flow plates for the dual cooled fuel assembly

MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

International Nuclear Information System (INIS)

Relloso, J.M.

1990-01-01

This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

CarbonNanoTubes (CNT) in bipolar plates for PEM fuel cell applications

Energy Technology Data Exchange (ETDEWEB)

Grundler, M.; Derieth, T.; Beckhaus, P.; Heinzel, A. [centre for fuel cell technology ZBT GmbH (Germany)

2010-07-01

Using standard mass production techniques for the fabrication of fuel cell components, such as bipolar plates, is a main issue for the commercialisation of PEM fuel cell systems. Bipolar plates contribute significantly to the cost structure of PEM stacks. In an upcoming fuel cell market a large number of bipolar plates with specific high-quality standards will be needed. At the Centre for Fuel Cell Technology (ZBT) together with the University of Duisburg-Essen fuel cell stacks based on injection moulded bipolar plates have been developed and demonstrated successfully [1]. This paper focuses on the interactions between carbon filling materials (graphite, carbon black and carbon nanotubes (CNT)) in compound based bipolar plates and especially the potential of CNTs, which were used in bipolar plates for the first time. The entire value added chain based on the feedstock, the compounding and injection moulding process, the component bipolar plate, up to the operation of a PEM single fuel cell stack with CNT-based bipolar plates is disclosed. (orig.)

Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

Energy Technology Data Exchange (ETDEWEB)

Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

2011-07-01

The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

Evaluation of plate type fuel options for small power reactors

International Nuclear Information System (INIS)

Andrzejewski, Claudio de Sa

2005-01-01

Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.

1985-01-01

Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235 U. Fuel plates containing 33 v/o U 3 Si and U 3 Si 2 behaved very well up to this burnup. Plates containing 33 v/o U 3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U 3 Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U 3 Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

PEM fuel cell bipolar plate material requirements for transportation applications

Energy Technology Data Exchange (ETDEWEB)

Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States)] [and others

1996-04-01

Cost effective bipolar plates are currently under development to help make proton exchange membrane (PEM) fuel cells commercially viable. Bipolar plates separate individual cells of the fuel cell stack, and thus must supply strength, be electrically conductive, provide for thermal control of the fuel stack, be a non-porous materials separating hydrogen and oxygen feed streams, be corrosion resistant, provide gas distribution for the feed streams and meet fuel stack cost targets. Candidate materials include conductive polymers and metal plates with corrosion resistant coatings. Possible metals include aluminium, titanium, iron/stainless steel and nickel.

SEM Characterization of an Irradiated Monolithic U-10Mo Fuel Plate

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.

2010-01-01

Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/cladding interface, particularly the interaction zone that had developed during fabrication and irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-Mo powders of an irradiated dispersion fuel.

Parametric study of the deformation of dispersion fuel plates

International Nuclear Information System (INIS)

Vieira, Edeval; Leal Neto, Ricardo Mendes; Durazzo, Michelangelo

2011-01-01

The Nuclear and Energy Research Institute - IPEN-CNEN/SP produces routinely the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U 3 Si 2 -Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed with base on information obtained from literature, which were used as premises for defining the current manufacturing procedures, according to a methodology with essentially empirical character. Despite the current rolling process to be perfectly stable and highly reproducible, it is not well characterized and therefore is not fully known. The objective of this work is to characterize the rolling process for producing fuel plates, presenting results of the evolution of all parameters of technological interest, after each rolling pass, obtaining information along the fuel plate deformation during the rolling process. (author)

Dry storage of MTR spent fuel from the Argentine radioisotope production reactor RA-3; Proyecto de compactado y reubicacion de los elementos combustibles quemados del RA-3 en el deposito de combustibles MTR del Centro Atomico Ezeiza

Energy Technology Data Exchange (ETDEWEB)

Di Marco, A; Gillaume, E J; Ruggirello, G; Zaweruchi, A [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Combustibles Nucleares

1997-12-31

The nuclear fuel elements of the RA-3 reactor consist in 19 rectangular fuel plates held in position by two lateral structural plates. The whole assembly is coupled to the lower nozzles that fits in the reactor core grid. The inner plates are 1.5 mm thick, 70.5 mm wide and 655 mm long and the outer plates are 100 mm longer. The fuel plates are formed by a core of an AI-U alloy co-laminated between two plates of Al. Enrichment is 90% {sup 235}U. After being extracted from the reactor, the fuel elements have been let to cool down in the reactor storage pool and finally moved to the storage facility. This facility is a grid of vertical underground channels connected by a piping system. The system is filled with processed and controlled water. At the present the storage capacity of the facility is near to be depleted and some indications of deterioration of the fuel elements has been detected. Due to the present status of the facility and the spent fuel stored there, a decision has been taken to proceed to modify the present underwater storage to dry storage. The project consist in: a) Decontamination and conditioning of the storage channels to prepare them for dry storage. b) Disassembly of the fuel elements in hot cells in order to can only the active fuel plates in an adequate tight canister. c) The remnant structural pieces will be treated as low level waste. (author). 10 figs.

Irradiation behavior of experimental miniature uranium silicide fuel plates

International Nuclear Information System (INIS)

Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

1983-01-01

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

Studies of mixed HEU-LEU-MTR cores using 3D models

Energy Technology Data Exchange (ETDEWEB)

Haenggi, P.; Lehmann, E.; Hammer, J.; Christen, R. [Paul Scherrer Institute, Villigen (Switzerland)

1997-08-01

Several different core loadings were assembled at the SAPHIR research reactor in Switzerland combining the available types of MTR-type fuel elements, consisting mainly of both HEU and LEU fuel. Bearing in mind the well known problems which can occur in such configurations (especially power peaking), investigations have been carried out for each new loading with a 2D neutron transport code (BOXER). The axial effects were approximated by a global buckling value and therefore the radial effects could be studied in considerably detail. Some of the results were reported at earlier RERTR meetings and were compared to those obtained by other methods and with experimental values. For the explicit study of the third dimension of the core, another code (SILWER), which has been developed in PSI for LWR power plant cores, has been selected. With the help of an adapted model for the MTR-core of SAPHIR, several important questions have been addressed. Among other aspects, the estimation of the axial contribution to the hot channel factors, the influence of the control rod position and of the Xe-poisoning on the power distribution were studied. Special attention was given to a core position where a new element was assumed placed near a empty, water filled position. The comparison of elements of low and high enrichments at this position was made in terms of the induced power peaks, with explicit consideration of axial effects. The program SILWER has proven to be applicable to MTR-cores for the investigation of axial effects. For routine use as for the support of reactor operation, this 3D code is a good supplement to the standard 2D model.

SEM characterization of an irradiated monolithic U-10Mo fuel plate

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.; Finlay, M.R.

2010-01-01

Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/AA6061 cladding interface, particularly the interaction zone that had developed during fabrication and any continued development during irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-7Mo powders of an irradiated dispersion fuel. (author)

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

Proposition of innovative and safe design of grid plate for Tehran research reactor

International Nuclear Information System (INIS)

Jalali, H.R.; Fadaei, A.H.

2017-01-01

Highlights: • An innovative and safe design for grid plate in research reactors proposed. • New grid plate acts as an independent shutdown system. • Neutronic and transient calculation was done using MTR-PC package. • Calculations show that the performance and safety of new design are acceptable. - Abstract: The purpose of this paper is to propose an innovative and safe design of grid plate for Tehran research reactor (TRR) without any reduction in its performance in comparison with the current operation. The new grid plate consisted of two joined cubic with empty walls which are place of fuels and heavy water, respectively. The proposed design is such that the reactor core is divided into two distinct parts using the heavy water. The heavy water is inserted in the walls of the new grid plate. The new design of grid plate by keeping the characteristics of the previous version creates the possibility of shutting the reactor down in critical condition. In this paper, at initial step, a simulation of acceptable benchmark for Tehran research reactor is performed which could be considered reliable and comparable with SAR (Safety Analysis Report) data. In the next step, two different designs are proposed for grid plate and then are applied to reactor core using simulation tools. For the proposed design: core excess reactivity, shutdown margin, control rod worth, neutron flux and kinetic parameters are calculated. Furthermore, the transient analysis was performed for the new design to check the status of reactor safety. Obtained results show that all neutronic parameters for the first operating core and the new design are comparable, and there is no reduction in the efficiency of reference core. Moreover, in the current design, a diverse and independent shutdown system for TRR was included. Nuclear reactor analysis codes including MTR-PC package were employed to carry out these calculations.

Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods

International Nuclear Information System (INIS)

Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul

2013-01-01

In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX

Ni-based amorphous alloy-coating for bipolar plate of PEM fuel cell by electrochemical plating

International Nuclear Information System (INIS)

Yamaura, S; Kim, S C; Inoue, A

2013-01-01

In this study, the Ni-Cr-P amorphous alloy-coated bipolar plates were produced by electro-plating on the Cu base plates with a flow field. The power generation tests of a single fuel cell with those Ni-Cr-P bipolar plates were conducted at 353 K. It was found that the single fuel cell with those Ni-Cr-P bipolar plates showed excellent I-V performance as well as that with the carbon graphite bipolar plates. It was also found that the single cell with those Ni-Cr-P bipolar plates showed better I-V performance than that with the Ni-P amorphous alloy-coated bipolar plates. Furthermore, the long-time operation test was conducted for 440 h with those Ni-Cr-P bipolar plates at the constant current density of 200 mA·cm −2 . As a result, it was found that the cell voltage gradually decreased at the beginning of the measurement before 300 h and then the voltage was kept constant after 300 h.

Prevention of criticality accidents. Fuel elements storage; Prevencion de accidentes de criticidad. Almacenamiento de elementos combustibles

Energy Technology Data Exchange (ETDEWEB)

Canavese, S I; Capadona, N M

1991-12-31

Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author). [Espanol] Partiendo de la necesidad de almacenar elementos combustibles tipo placa MTR (Materials Testing Reactors), producidos con uranio enriquecido al 20% en U235 para reactores de investigacion, se requiere el diseno de un deposito para tal fin que brinde esencialmente un alto grado de seguridad intrinseca y que no ofrezca complicaciones en cuanto a su construccion. (Autor).

Irradiation testing of miniature fuel plates for the RERTR program

Energy Technology Data Exchange (ETDEWEB)

Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

1983-08-01

An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

Note on current position regarding the development by the UKAEA of Reduced Enrichment fuels for Research and Test Reactors

International Nuclear Information System (INIS)

Hickey, B.

1983-01-01

The United Kingdom Atomic Energy Authority have an MTR fuel fabrication plant located at Dounreay on the north coast of Scotland. The prime function of the plant is to manufacture fuel elements for the UKAEA's own DIDO and PLUTO heavy water reactors located at their research establishment at Harwell. The plant, which has a capacity of about 1000 fuel elements per annum, also manufactures fuel elements, on a commercial basis, for university reactors in the United Kingdom and for a number of customers in overseas countries. The UKAEA have been manufacturing MTR fuel elements of a wide range of designs for over twenty-five years. Following the initiative of the US Government's RERTR programme, the UKAEA have embarked on a modest programme of MTR fuel manufacturing development., irradiation and post-irradiation examination to establish the techniques required to manufacture fuel elements containing uranium of a significantly lower enrichment than that in the fuel elements they currently manufacture. In the first instance this work is being directed towards the production of fuel elements containing uranium of 45% enrichment. After an initial analysis it was recognised that although a satisfactory 45% enriched version of certain of the designs of fuel elements currently manufactured could probably be produced using established U/Al alloy technology, it would be necessary to utilise powder technology for other elements in order to achieve the higher uranium density required. Studies of published information and consideration of the technology and facilities already available at Dounreay prompted the decision to concentrate on the development Of U 3 O 8 /Al cermet type fuel elements of similar geometry to those currently manufactured. Some of the fuel element designs currently manufactured by the UKAEA are listed: Concentric (Extruded) 74% enriched; Concentric Plates 80% enriched with densities 0.60 and 0.53 g U/ cm 3 ; Flat Plate (Swaged) 80% enriched and Flat Plate

Computer simulation of variform fuel assemblies using Dragon code

International Nuclear Information System (INIS)

Ju Haitao; Wu Hongchun; Yao Dong

2005-01-01

The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

Caramel, uranium oxide fuel plates for water cooled reactors

International Nuclear Information System (INIS)

Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

1979-01-01

The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

Long term immersion test of aluminum alloy AA 6061 used for fuel cladding in MTR type reactors

International Nuclear Information System (INIS)

Linardi, Evelina M.; Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana

2009-01-01

In this work we present the results of long term immersion tests performed in the aluminum alloy AA 6061, used for fuel cladding in MTR type reactors. The tests were performed at open circuit potential in high purity water (ρ = 18.2 MΩ.cm) and in 10 -3 M NaCl solution. Two kinds of assemblies were studied: simple sheets and artificial crevices, immersed during 6, 12 and 18 months at room temperature. In both media and both assemblies, the aluminum hydroxide phases crystalline bayerite and bohemite were identified. It was found that a kind of localized attack named alkaline attack occurs around the iron-rich intermetallics. These particles were confirmed to control the corrosion of the AA 6061 alloy in an aerated medium. Immersion times for up to 18 months did not increase the oxide growth or the alkaline attack on the AA 6061 alloy. (author)

Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1991-10-01

Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

Back-end of the research reactor fuel cycle

International Nuclear Information System (INIS)

Gruber, Gehard J.

1996-01-01

This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (author)

The Transcriptional Repressor, MtrR, of the mtrCDE Efflux Pump Operon of Neisseria gonorrhoeae Can Also Serve as an Activator of “off Target” Gene (glnE Expression

Directory of Open Access Journals (Sweden)

Paul J. T. Johnson

2015-06-01

Full Text Available MtrR is a well-characterized repressor of the Neisseria gonorrhoeae mtrCDE efflux pump operon. However, results from a previous transcriptional profiling study suggested that MtrR also represses or activates expression of at least sixty genes outside of the mtr locus. Evidence that MtrR can directly repress so-called “off target” genes has previously been reported; in particular, MtrR was shown to directly repress glnA, which encodes glutamine synthetase. In contrast, evidence for the ability of MtrR to directly activate expression of gonococcal genes has been lacking; herein, we provide such evidence. We now report that MtrR has the ability to directly activate expression of glnE, which encodes the dual functional adenyltransferase/deadenylase enzyme GlnE that modifies GlnA resulting in regulation of its role in glutamine biosynthesis. With its capacity to repress expression of glnA, the results presented herein emphasize the diverse and often opposing regulatory properties of MtrR that likely contributes to the overall physiology and metabolism of N. gonorrhoeae.

Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U3Si2 in RECH-1

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J.

1999-01-01

The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U 3 Si 2 -Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (author)

Characterisation of the corrosion products of non-irradiated material test reactors fuel elements (MTR-FE)

Energy Technology Data Exchange (ETDEWEB)

Mazeina, L.; Curtius, H.; Fachinger, J. [Inst. for Safety Research and Reactor Technology, Research Centre Juelich (Germany)

2003-07-01

In a high concentrated Mg-rich brine a non-irradiated MTR-FE corroded. The formed corrosion products consists of an amorphous part and of hydrotalcites, which were identified as Mg-Al-hydrotalcites with chloride anions in the interlayer. (orig.)

Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

2011-01-01

Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

Analysis of hydraulic instability of ANS involute fuel plates

International Nuclear Information System (INIS)

Sartory, W.K.

1991-11-01

Curved shell equations for the involute Advanced Neutron Source (ANS) fuel plates are coupled to two-dimensional hydraulic channel flow equations that include fluid friction. A complete set of fluid and plate boundary conditions is applied at the entrance and exit and along the sides of the plate and the channel. The coupled system is linearized and solved to assess the hydraulic instability of the plates

Determination of the burn-up in fuels of the MTR type by means of gamma spectroscopy with crystal of INa(Tl)

International Nuclear Information System (INIS)

Kestelman, A.J.

1988-01-01

One of the responsibilities of the Laboratory of Analysis by Neutronic Activation of the RA-6 reactor is to determine the burn-up in fuels of the MTR type. In order to gain experience, up to the arrival of the hyperpure Germanium detector (HPGe) to be used in normal operation, preliminary measurements with a crystal of INa(Tl) were made. The fuel elements used are originated in the RA-3 reactor, with a decay superior to the thirteen years. For this reason, the unique visible photoelectric peak is the one of Cs-137, owing to the low resolution of the INa(Tl). After preliminary measurements, the profiles of burn-up, rectified by attenuation, were measured. Once the efficiency of the detector was determined, the calculation of the burn-up was made; for the element No. 144, a value of 21.6 ± 2.9 g was obtained to be compared with the value 21.9 g which was the evaluation made by the operators. (Author) [es

Characterization of an irradiated RERTR-7 fuel plate using transmission electron microscopy

International Nuclear Information System (INIS)

Gan, J.; Keiser, D.D. Jr.; Miller, B.D.; Robinson, A.B.; Medvedev, P.

2010-01-01

Transmission electron microscopy (TEM) has been used to characterize an irradiated fuel plate with Al-2Si matrix from the Reduced Enrichment Research and Test Reactor RERTR-7 experiment that was irradiated under moderate reactor conditions. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behaviour during irradiation. Furthermore, TEM analysis showed that the Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation. An important question that remains to be answered about the irradiation behaviour of U-Mo dispersion fuels is how do more aggressive irradiation conditions affect the behaviour of fission gases within the U-7Mo fuel particles and in the amorphous interaction layers on the microstructural scale that can be characterized using TEM? This paper will discuss the results of TEM analysis that was performed on a sample taken from an irradiated RERTR-7 fuel plate with Al-2Si matrix. This plate was exposed to more aggressive irradiation conditions than the RERTR-6 plate. The microstructural features present within the U-7Mo and the amorphous interaction layers will be discussed. The results of this analysis will be compared to what was observed in the earlier RERTR-6 fuel plate characterization. (author)

Numerical simulation research on rolling process of monolithic nuclear fuel plate

International Nuclear Information System (INIS)

Wan Jibo; Kong Xiangzhe; Ding Shurong; Xu Hongbin; Huo Yongzhong

2015-01-01

For the strain-rate-dependent constitutive relation of zircaloy cladding in UMo monolithic nuclear fuel plates, the three-dimensional stress updating algorithm was derived out, and the corresponding VUMAT subroutine to define its constitutive relation was developed and validated; the finite element model was built to simulate the frame rolling process of UMo monolithic nuclear fuel plates; with the explicit dynamic finite element method, the evolution rules of the deformation and contact pressure during the rolling process within the composite slab were obtained and analyzed. The research results indicate that it is convenient and efficient to define the strain-rate- dependent constitutive relations of materials with the user-defined material subroutine VUMAT; the rolling-induced contact pressure between the fuel meat and the covers varies with time, and the maximum pressure exits at the symmetric plane along the plate width direction. This study supplies a foundation and a computation method for optimizing the processing parameters to manufacture UMo monolithic nuclear fuel plates. (authors)

Fuel plate stability experiments and analysis for the Advanced Neutron Source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

1992-01-01

The planned Advanced Neutron Source (ANS) and several existing reactors use closely spaced arrays of involute shaped fuel-plates which are cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported in this paper. The tests were conducted using full scale epoxy plate models of the aluminum/uranium silicide ANS involute shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as function of the flow velocity are examined. Comparisons with mathematical models are noted. 12 refs

Fuel plate stability experiments and analysis for the Advanced Neutron Source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

1993-05-01

The planned reactor for the Advanced Neutron Source (ANS) will use closely spaced arrays of involute-shaped fuel plates that will be cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities, adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported. The tests were conducted using full-scale epoxy plate models of the aluminum/uranium silicide ANS involute-shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as functions of the flow velocity are examined. Comparisons with mathematical models are noted

Automated ultrasonic scanning of flat plate nuclear fuel

International Nuclear Information System (INIS)

Barna, B.A.

1979-01-01

One of the most challenging problems in Non-Destructive Testing lies in making the inspection as rapid, precise, cost effective and operator independent as possible. Only by optimizing these four factors can a technology take full advantage of the quality control possible with NDT. This paper describes a highly complex application of high frequency ultrasonics to image extremely small and difficult to detect flaws in a production line environment. The objects of interest are flat plate nuclear fuel used in the Advanced Test Reactor at the Idaho National Engineering Laboratory. The plates are fabricated by hot rolling a sandwich of alloyed uranium fuel and aluminum cladding. After rolling, the block is flattened to a long thin plate approximately 1.27 m (55 inches) long, 102 mm (4 inches) wide and 1.25 mm (0.050 inches) thick. The core, or fuel area is nominally 0.75 mm (0.030 inches) thick with 0.25 mm (0.010 inches) of aluminum bonded to both sides. As might be expected the fabrication is a sensitive process which can introduce several flaws detrimental to the reactor operation if they are undetected. Two of the characteristics that must be examined are the cladding thickness of the aluminum left over the fuel and the quality of bond between the cladding and the fuel. If either the cladding is too thin or the bonding inadequate thermal and/or corrosive activity can crack the protective cladding

Highly conductive composites for fuel cell flow field plates and bipolar plates

Science.gov (United States)

Jang, Bor Z; Zhamu, Aruna; Song, Lulu

2014-10-21

This invention provides a fuel cell flow field plate or bipolar plate having flow channels on faces of the plate, comprising an electrically conductive polymer composite. The composite is composed of (A) at least 50% by weight of a conductive filler, comprising at least 5% by weight reinforcement fibers, expanded graphite platelets, graphitic nano-fibers, and/or carbon nano-tubes; (B) polymer matrix material at 1 to 49.9% by weight; and (C) a polymer binder at 0.1 to 10% by weight; wherein the sum of the conductive filler weight %, polymer matrix weight % and polymer binder weight % equals 100% and the bulk electrical conductivity of the flow field or bipolar plate is at least 100 S/cm. The invention also provides a continuous process for cost-effective mass production of the conductive composite-based flow field or bipolar plate.

Development of core technology for research reactors using plate type fuels

International Nuclear Information System (INIS)

Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

2009-12-01

Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

International Nuclear Information System (INIS)

Tolosa, S.C.; Marajofsky, A.

2004-01-01

In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate

An expert system to analyze homogeneity in fuel element plates for research reactors

International Nuclear Information System (INIS)

Cativa Tolosa, Sebastian; Marajofsky, Adolfo

2004-01-01

In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up.This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to X-ray images. These images are generated when the X-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized X-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate. (author)

Alternative bipolar plates design and manufacturing for PEM fuel cell

International Nuclear Information System (INIS)

Lee Chang Chuan; Norhamidi Muhamad; Jaafar Sahari

2006-01-01

Bipolar plates is one of the important components in fuel cell stack, it comprise up to 80% of the stack volume. Traditionally, these plates have been fabricated from graphite, owing to its chemical nobility, and high electrical and thermal conductivity; but these plates are brittle and relatively thick. Therefore increasing the stack volume and size. Alternatives to graphite are carbon-carbon composite, carbon-polymer composite and metal (aluminum, stainless steel, titanium and nickel based alloy). The use of coated and uncoated metal bipolar plates has received attention recently due to the simplicity of plate manufacturing. The thin nature of the metal substrate allows for smaller stack design with reduced weight. Lightweight coated metals as alternative to graphite plate is being developed. Beside the traditional method of machining and slurry molding, metal foam for bipolar plates fabrication seems to be a good alternative. The plates will be produced with titanium powder by Powder Metallurgy method using space holders technique to produce the meal foam flow-field. This work intends to facilitate the materials and manufacturing process requirements to produce cost effective foamed bipolar plates for fuel cell

Application of MTR soft-decision decoding in multiple-head ...

Indian Academy of Sciences (India)

basic MTR logic circuits, and to develop, a new one, the soft-decision MTR decoder, based on such ... of integrated circuits provides their quite simple realization. ..... recording channels, PSU-UNS International Conference on Engineering and ...

SEM and TEM Characterization of As-Fabricated U-7Mo Disperson Fuel Plates

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Yao, B.; Perez, E.; Sohn, Y.H.

2009-01-01

The starting microstructure of a dispersion fuel plate can have a dramatic impact on the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of dispersion fuel plates, SEM and TEM analysis have been performed on RERTR-9A archive fuel plates, which went through an additional hot isostatic procsssing (HIP) step during fabrication. The fuel plates had depleted U-7Mo fuel particles dispersed in either Al-2Si or 4043 Al alloy matrix. For the characterized samples, it was observed that a large fraction of the ?-phase U-7Mo alloy particles had decomposed during fabrication, and in areas near the fuel/matrix interface where the transformation products were present significant fuel/matrix interaction had occurred. Relatively thin Si-rich interaction layers were also observed around the U-7Mo particles. In the thick interaction layers, (U)(Al,Si)3 and U6Mo4Al43 were identified, and in the thin interaction layers U(Al,Si)3, U3Si3Al2, U3Si5, and USi1.88-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this work, exposure of dispersion fuel plates to relatively high temperatures during fabrication impacts the overall microstructure, particularly the nature of the interaction layers around the fuel particles. The time and temperature of fabrication should be carefully controlled in order to produce the most uniform Si-rich layers around the U-7Mo particles.

Composite Bipolar Plate for Unitized Fuel Cell/Electrolyzer Systems

Science.gov (United States)

Mittelsteadt, Cortney K.; Braff, William

2009-01-01

In a substantial improvement over present alkaline systems, an advanced hybrid bipolar plate for a unitized fuel cell/electrolyzer has been developed. This design, which operates on pure feed streams (H2/O2 and water, respectively) consists of a porous metallic foil filled with a polymer that has very high water transport properties. Combined with a second metallic plate, the pore-filled metallic plates form a bipolar plate with an empty cavity in the center.

Qualification of high-density fuel manufacturing for research reactors at CNEA

Energy Technology Data Exchange (ETDEWEB)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [CNEA, Buenos Aires (Argentina)

2001-07-01

CNEA, the National Atomic Energy Commission of Argentina, is at the present a qualified supplier of uranium oxide fuel for research reactors. A new objective in this field is to develop and qualify the manufacturing of LEU high-density fuel for this type of reactors. According with the international trend Silicide fuel and U-xMo fuel are included in our program as the most suitable options. The facilities to complete the qualification of high-density MTR fuels, like the manufacturing plant installations, the reactor, the pool side fuel examination station and the hot cells are fully operational and equipped to perform all the activities required within the program. The programs for both type of fuels include similar activities: development and set up of the fuel material manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of miniplates, fabrication and irradiation of full scale fuel elements, post-irradiation examination and feedback for manufacturing improvements. For silicide fuels most of these steps have already been completed. For U-xMo fuel the activities also include the development of alternative ways to obtain U-xMo powder, feasibility studies for large-scale manufacturing and the economical assessment. Set up of U-xMo fuel plate manufacturing is also well advanced and the fabrication of the first full scale prototype is foreseen during this year. (author)

ANALISIS POLA MANAJEMEN BAHAN BAKAR DESAIN TERAS REAKTOR RISET TIPE MTR

Directory of Open Access Journals (Sweden)

Lily Suparlina

2015-03-01

Full Text Available Parameter neutronik dibutuhkan dalam mendesain teras reaktor riset. Reaktor riset jenis MTR (Material Testing Reactor sangat diminati karena dapat digunakan baik untuk riset dan juga produksi radio isotop. Reaktor riset yang ada saat ini sudah tua sehingga dibutuhkan desain reaktor yang mempunyai teras kompak. Desain teras reaktor riset yang sudah ada saat ini belum cukup memadai untuk memenuhi persyaratan di dalam UCD yang telah ditetapkan yaitu fluks neutron termal di teras 1x1015 n/cm2s, oleh karena itu perlu dibuat desain teras reaktor baru sebagai alternatif yang kompak dan dapat menghasilkan fluks neutron tinggi. Telah dilakukan perhitungan dan analisis terhadap manajemen bahan bakar desain teras kompak dengan konfigurasi teras 5x5, berbahan bakar U9Mo-Al dan tinggi teras aktif 70 cm. Tujuan dari riset ini untuk memperoleh fluks neutron di teras memenuhi kebutuhan seperti yang telah ditetapkan di UCD dengan panjang siklus operasi minimum 20 hari pada daya 50 MW. Perhitungan dilakukan dengan menggunakan paket program komputer WIMSD-5B untuk menggenerasi tampang lintang makroskopik bahan bakar dan Batan-FUEL untuk memperoleh nilai parameter neutronik serta Batan-3DIFF untuk perhitungan nilai reaktivitas batang kendali. Perhitungan parameter neutronik teras reaktor riset ini dilakukan untuk bahan bakar U-9Mo-Al dengan tingkat muat bervariasi dan 2 macam pola pergantian bahan bakar yaitu teras segar dan teras setimbang. Hasil analisis menunjukkan bahwa pada teras segar, tingkat muat 235U sebesar 360 gram, 390 gram dan 450 gram memenuhi kriteria keselamatan dan kriteria penerimaan di UCD dengan nilai fluks neutron termal di teras lebih dari 1x1015 n/cm2s dan panjang siklus >20 hari, sedangkan pada teras setimbang panjang siklus dapat terpenuhi hanya untuk tingkat muat 450 gram. Kata kunci: desain teras reaktor, bahan bakar UMo, pola bahan bakar, WIMS, BATAN-FUEL   Research reactor core design needs neutronics parameter calculation use computer

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

2009-08-01

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

International Nuclear Information System (INIS)

Travelli, A.

1988-01-01

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

Release of fission products from miniature fuel plates at elevated temperature

International Nuclear Information System (INIS)

Posey, J.C.

1982-01-01

Three miniature fuel plates were tested at progressively higher temperatures. A U 3 Si plated blistered and released fission gases at 500 0 C. Two U 3 O 8 filled plates blistered and released fission gases at 550 0 C

Reprocessing of LEU silicide fuel at Dounreay

International Nuclear Information System (INIS)

Cartwright, P.

1996-01-01

UKAEA have recently reprocessed two LEU silicide fuel elements in their MTR fuel reprocessing plant at Dounreay. The reprocessing was undertaken to demonstrate UKAEA's commitment to the world-wide research reactor communities future needs. Reprocessing of LEU silicide fuel is seen as a waste treatment process, resulting in the production of a liquid feed suitable for conditioning in a stable form of disposal. The uranium product from the reprocessing can be used as a blending feed with the HEU to produce LEU for use in the MTR cycle. (author)

Design of fuel element for RA10

International Nuclear Information System (INIS)

Estevez, Esteban A.; Markiewicz, Mario; Gerding, Roberto

2012-01-01

The RA-10 reactor is an open pool multipurpose reactor. It is intended for radioisotopes production, fuel irradiation and use of neutron beam experiments. The nominal configuration core consists of 19 fuel elements (FE) and 6 in-core irradiation positions. With regard to the FE, although both conceptual design and manufacturing technology are similar to the already developed and qualified by CNEA (MTR fuel flat plate), the conditions imposed by the new reactor on FE's are more demanding that previous supplies. Here it should be mentioned the magnitude of the hydrodynamic forces acting on the FE caused by coolant flow through the core (upward) and mainly by the high coolant velocity between fuel plates (greater than 5 times than those currently in operation). Moreover, the high power density results in higher heat flux in fuel plates and greater temperature gradient. As a result of these increased demands present during irradiation, and in order to maintain a high level of reliability, it is necessary carry out some modifications in the mechanical design of the FE (with respect to the so-called ECBE design or s tandard ) . Design verification is performed through analytical and code calculations, and hydrodynamic tests on a full-scale prototype. This article describes the design of the FE for RA 10 reactor, with special emphasis on those aspects that represent innovations in the traditional design (ECBE). It also presents the functional requirements, design criteria and design limits established according to the reactor operational states (author)

Release of fission products from miniature fuel plates at elevated temperature

International Nuclear Information System (INIS)

Posey, John C.

1983-01-01

Three miniature fuel plates were tested at progressively higher temperatures. A U 3 Si filled plate blistered and released fission gases at 500 deg. C. Two U 3 O 8 filled plates blistered and released fission gases at 550 deg. C. (author)

Progress in the development of uranium silicide (U3Si2) fuel at BATAN

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1995-01-01

After successful fabrication of two full-size prototype fuel elements containing ∼3.0 gU/cm 3 in the form of U 3 Si 2 -Al dispersion now undergoing irradiation in the Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) core since 1990, further development in U 3 Si 2 -A2 dispersion fuel element manufacturing has been pursued, whose progress in discussed in this paper, with a special attention on the use of much higher-loading aimed at obtaining a better understanding on the influence of higher-loading on fuel core and plate manufacturing and quality. At present, high-loading U 3 Si 2 -AI dispersion miniplates are being manufactured for preparing some mini-fuel elements to be test-irradiated in the new MTR in-pile loop of the RSG-GAS. (author)

Parametric study of the deformation of U3Si2-Al dispersion fuel plates

International Nuclear Information System (INIS)

Vieira, Edeval

2011-01-01

The Nuclear and Energy Research Institute - IPEN-CNEN/SP produces routinely the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U 3 Si 2 -Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed with base on information obtained from literature, which were used as premises for defining the current manufacturing procedures, according to a methodology with essentially empirical character. Despite the current rolling process to be perfectly stable and highly reproducible, it is not well characterized and therefore is not fully known. The objective of this work is to characterize the rolling process for producing fuel plates, specifically the evolution of dimensional parameters of the fuel plate as a function of its deformation in the rolling process. Results are presented in terms of the evolution of the thickness of the fuel meat and cladding of the fuel plate along the deformation, as well as the terminals defects, microstructure and porosity of the fuel meat. (author)

Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Lee, Yoon Sang; Cheong, Yong Moo

2010-01-01

Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

International Nuclear Information System (INIS)

Jo, Daeseong; Kim, Haksung

2015-01-01

Highlights: • Neutronic and thermal-hydraulic analyses of U–Mo fuel irradiated in HANARO reactor. • A mock-up irradiation target was designed and tested to measure the flow rate. • During normal operation, boiling does not occur. • During limiting accidents, boiling occurs. However, fuel integrity is maintained. - Abstract: Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2

A model development for a thermohydraulic calculation material convection of MTR (Materials Testing Reactors)

International Nuclear Information System (INIS)

Abbate, P.

1990-01-01

The CONVEC program developed for the thermohydraulic calculation under a natural convection regime for MTR type reactors is presented. The program is based on a stationary, one dimensional model of finite differences that allow to calculate the temperatures of cooler, cladding and fuel as well as the flow for a power level specified by the user. This model has been satisfactorily validated by a water cooling (liquid phase) and air system. (Author) [es

Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel

International Nuclear Information System (INIS)

Todd, Lawrence E.; Pace, Brett W.

1996-01-01

This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)

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.

Connection between end plates and rods in a BWR fuel element

International Nuclear Information System (INIS)

Cali', G.P.

1975-01-01

The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

Parametric study of fission-induced U-Mo fuel creep and structural analysis of fuel plates in view of implications for microstructure evolution

International Nuclear Information System (INIS)

Kim, Y.S.; Hofman, G.L.; Choo, Y.S.; Robinson, A.B.

2010-01-01

U-Mo fuel deformation during irradiation in U-Mo/Al dispersion plates is investigated by using the irradiation data from the RERTR-3 through -9 tests. The observation of fuel particle sintering during irradiation is also presented and its influence for fuel performance is discussed. Structural analysis was also performed to examine the relationship between the stress distribution in the plate and the location of matrix-pore formation in the plate. (author)

The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

International Nuclear Information System (INIS)

Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

2009-01-01

The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

International Nuclear Information System (INIS)

Muellauer, J.

1989-01-01

Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

Immobilisation of MTR waste in cement (product evaluation). Final report. December 1987

International Nuclear Information System (INIS)

Howard, C.G.; Lee, D.J.

1988-01-01

The enriched uranium/aluminium fuel used in Material Testing Reactors is reprocessed at Dounreay Nuclear Power Development Establishment (DNE). The main chemical component of the liquid waste produced by this process is acid deficient aluminium nitrate. This is stored in stainless steel tanks at DNE. As a result of work carried out under the UKAEA radioactive waste management programme a decision was taken to immobilise the waste in cement. The programme had two main components, plant design and development of the cementation process. The plant for the cementation of MTR waste is under construction and will be commissioned in 1988/9. The primary objective of this project is to find a suitable process for changing the highly mobile radioactive waste into an inert stable solid. Work carried out on the development of the immobilisation process showed that a conditioning stage (neutralisation) is required to make the acid waste compatible with cement. Small scale experiments showed that adding Ordinary Portland Cement blended with ground granulated Blast Furnace Slag to Simulant MTR Liquor produces an acceptable product. The process has been demonstrated at full scale (200 litres) and the products have been subjected to an extensive programme of destructive and non-destructive testing. Specimens have been tested up to 1200 days after manufacture and show no significant signs of deterioration even when stored underwater or when subjected to freeze thaw cycling. Development work has also shown that the process can successfully immobilise simulant MTR liquor over a wide range of liquor concentrations. The programme therefore successfully produced a formulation that met all the requirements of both the process and product specification. (author)

Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.

1998-01-01

The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their 137 Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the 137 Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A 137 Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment

Successful completion of a time sensitive MTR and TRIGA Indonesian shipment

International Nuclear Information System (INIS)

Anne, Catherine; Patterson, John; Messick, Chuck

2005-01-01

Early this year, a shipment of 109 MTR fuel assemblies was received at the Department of Energy's Savannah River Site from the BATAN reactor in Serpong, Indonesia and another of 181 TRIGA fuel assemblies was received at the Idaho National Laboratory from the two BATAN Indonesian TRIGA reactors in Bandung and Yogyakarta, Indonesia. These were the first Other-Than- High-Income Countries shipments under the FRR program since the Spring 2001. The Global Threat Reduction Initiative announced by Secretary Abraham will require expeditious scheduling and extreme sensitivity to shipment security. The subject shipments demonstrated exceptional performance in both respects. Indonesian terrorist acts and 9/11 impacted the security requirements for the spent nuclear fuel shipments. Internal Indonesian security issues and an upcoming Indonesian election led to a request to perform the shipment with a very short schedule. Preliminary site assessments were performed in November 2003. The DOE awarded a task order to NAC for shipment performance just before Christmas 2003. The casks departed the US in January and the fuel elements were delivered at the DOE sites by the end of April 2004. The paper will present how the team completed a successful shipment in a timely manner. (author)

Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

Energy Technology Data Exchange (ETDEWEB)

A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

2009-08-01

A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

Energy Technology Data Exchange (ETDEWEB)

Samuel J. Miller; Hakan Ozaltun

2012-11-01

This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

Fuel cell plates with skewed process channels for uniform distribution of stack compression load

Science.gov (United States)

Granata, Jr., Samuel J.; Woodle, Boyd M.

1989-01-01

An electrochemical fuel cell includes an anode electrode, a cathode electrode, an electrolyte matrix sandwiched between electrodes, and a pair of plates above and below the electrodes. The plate above the electrodes has a lower surface with a first group of process gas flow channels formed thereon and the plate below the electrodes has an upper surface with a second group of process gas flow channels formed thereon. The channels of each group extend generally parallel to one another. The improvement comprises the process gas flow channels on the lower surface of the plate above the anode electrode and the process gas flow channels on the upper surface of the plate below the cathode electrode being skewed in opposite directions such that contact areas of the surfaces of the plates through the electrodes are formed in crisscross arrangements. Also, the plates have at least one groove in areas of the surfaces thereof where the channels are absent for holding process gas and increasing electrochemical activity of the fuel cell. The groove in each plate surface intersects with the process channels therein. Also, the opposite surfaces of a bipolar plate for a fuel cell contain first and second arrangements of process gas flow channels in the respective surfaces which are skewed the same amount in opposite directions relative to the longitudinal centerline of the plate.

Immobilisation of MTR waste in cement (product evaluation)

International Nuclear Information System (INIS)

Howard, C.G.; Lee, D.J.

1988-01-01

The enriched uranium/aluminium fuel used in Material Testing Reactors is reprocessed at Dounreay Nuclear Power Development Establishment (DNE). The main chemical component of the liquid waste produced by this process is acid deficient aluminium nitrate. The primary objective of this project is to find a suitable process for changing the highly mobile radioactive waste into an inert stable solid. Work carried out on the development of the immobilisation process showed that a conditioning stage (neutralisation) is required to make the acid waste compatible with cement. Small scale experiments showed that adding Ordinary Portland Cement blended with ground granulated Blast Furnace Slag to Simulant MTR Liquor produces an acceptable product. The process has been demonstrated at full scale (200 litres) and the products have been subjected to an extensive programme of destructive and non-destructive testing. (author)

Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

International Nuclear Information System (INIS)

Zhang Zhuohua; Yu Junchong; Peng Shinian

2014-01-01

According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

Energy Technology Data Exchange (ETDEWEB)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Determination of elastic constants of fuels plates based on uranium by ultrasound testing

International Nuclear Information System (INIS)

Moreira Castro, Martin Ignacio

2015-01-01

Current nuclear reactors use as U-235 U-enriched compounds enriched with U-235, requiring U-alloys that increase the amount of atoms available for nuclear fission in a convenient way. This study was carried out on fuel plates manufactured in the Chilean Nuclear Energy Commission, whose cores are composed of a dispersed mixture Al-U_3Si_2 and Al-U_7Mo, with different densities of uranium, covered by a coating of Al6061. The objective was to characterize elastically and classify the fuel plates analyzed. Specifically, five Al-U_3Si_2 fuel plates with 1.7 gU/cm"3, eight A-U_3Si_2 with 3.4 gU/cm"3, five of A-l U_3Si_2 with 4.8 gU/cm"3 were successfully studied. The apparent elastic constants (Young and Shear modules, and Poisson coefficient) were determined in the area where the fuel is located (MEAT) by means of an ultrasound sampling technique, thus being able to characterize them and classify them according to their composition. The behavior of the elastic constants generally shows a tendency to decrease as the amount of U_3Si_2 particles dispersed in the MEAT zone of the fuel plates increases. In addition, the non-destructive test method used made it possible to detect several differences between the fuel plates analyzed, such as the amount of reduction in rolling, among others. Additionally, six experimental fuel miniplates were analyzed whose meat were formed by a dispersion of the Al-UMo type, specifically: two of Al-U_7Mo with 6.0 gU/cm"3, two of Al-U_7Mo with 7.0 gU/ cm"3 and two of Al-U_7Mo with 8.0 gU/cm"3. The response of the U-Mo fuel miniplates against this technique was not good, so several ideas were proposed to improve this situation

Criticality Studies in a Pilot Plant for Processing MTR-Type Irradiated Fuels; Estudios de Criticidad de una Planta Piloto para el Tratamiento de Combustibles Irradiados Tipo ' MTR '

Energy Technology Data Exchange (ETDEWEB)

Pereira Sanchez, G.; Uriarte Hueda, A. [Junta de Energia Nuclear, Division de Materiales Madrid (Spain)

1966-05-15

A number of theoretical studies on nuclear safety have been carried out in a pilot plant being constructed at the Junta de Energia Nuclear in Madrid for processing irradiated fuels from the MTR-type experimental reactor JEN-1. The study was carried out working with aqueous and organic solutions at two levels of {sup 235}U enrichment - 20% and 93%. The paper is divided into two main parts: the first deals with the individual items of equipment, and the interactions between these are studied in the second part. The calculations in this second part have been made using three different methods to make it more certain that the system as a whole can never be critical. The first method employed is based on the solid angle concept and makes it possible to fix the maximum {sup 235}U concentrations within the system. The second method, based on the albedo, supplies the value of the multiplication factor K of the whole assembly as a function of the concentration of {sup 235}U. In the last part, the distribution of the equipment is compared with other similar systems and experimental tests from other sources. Finally, the paper specifies the conditions for working the installation which ensure that a nuclear accident can never occur. (author) [Spanish] Se ha efectuado una serie de estudios teoricos sobre la seguridad nuclear de una planta piloto, que se encuentra en construccion en la Junta de Energfa Nuclear situada en Madrid, para el tratamiento de combustibles irradiados procedentes del reactor experimental JEN-1 del tipo MTR. El estudio se ha realizado utilizando disoluciones, tanto acuosas como organicas, con dos grados de enriquecimiento, 20% y 93% en {sup 235}U. Este trabajo comprende dos partes principales: en la primera se han considerado las distintas unidades del equipo individualmente y en la segunda se han estudiado las interacciones entre ellas. El calculo de esta segunda parte se ha hecho por tres metodos diferentes para tener una mayor seguridad de que el

Development of Fusion Nuclear Technologies and the role of MTR's

International Nuclear Information System (INIS)

Laan, J.G. van der; Schaaf, B. van der

2006-01-01

Fusion power plant operation will strongly depend on the economy and reliability of crucial components, such as first wall modules, tritium breeding blankets and divertors. Their operating temperature shall be high to accomplish high plant efficiency. The materials properties and component fabrication routes shall also assure long reliable operation to minimize plant outage. The components must be fabricated in large quantities based on demonstrations with a limited amount of test beds. Mock-ups and test loops will, through iteration processes, demonstrate the reliable operation under reference thermal-hydraulic conditions. Although 14 MeV neutrons dominate the nuclear conditions near the first wall, neutron transport analyses have shown that large portions of the components near the plasma have to cope with a neutron spectrum resembling a fission core. Present Materials Test Reactors, MTR's, offer fluxes relevant for large parts of the fusion major components. The mixed and fast fission spectra though is not representative for all fusion conditions. The strong point of MTR's is their ability to generate sufficient displacement damage in the materials in a relatively short time. The cores of MTR's provide sufficient space for irradiation of representative cut-outs of components to allow integrated functional and materials tests in a high flux neutron field. The MTR's are the primary test bed for structural and functional fusion relevant materials. The MTR space and dose rates provide a valuable base line for the developments and demonstrations of fusion key components in a neutron field. In recent years the pebble bed assembly, PBA, irradiated in the HFR, Petten, has shown the feasibility of the helium-cooled concept with lithium ceramics and beryllium multiplier pebble beds. The irradiations produce a wealth of process parameters for the control of the tritium release of the pebbles. The PBA packaging, cooling and tritium purging arrangements closely resemble the

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

Energy Technology Data Exchange (ETDEWEB)

Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-01-26

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

Energy Technology Data Exchange (ETDEWEB)

Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

2003-07-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

International interest in the BONAPARTE measurement bench. Post-irradiation examination of lower-enriched fuel plates

International Nuclear Information System (INIS)

2014-01-01

The Belgian Nuclear Research Center SCK-CEN has developed a measurement bench (BONAPARTE) for the non-destructive analysis on fuel plate and rod type fuel elements. BONAPARTE is a modular device that can be employed for many purposes. The article discusses the employment of the BONAPARTE device for the accurate full post-irradiation mapping of fuel plate swelling with degree of precision of just a few micrometers.

A review of microstructural analysis on U3Si2-Al plate-type fuel

International Nuclear Information System (INIS)

Ti Zhongxin; Guo Yibai

1995-12-01

The microstructure of U 3 Si 2 -Al plate-type fuel, that is the microstructure of fuel particles, compatibility of the fuel particles and Al matrix, fuel particles distribution, dogbone area morphology, clad and meat thickness, bone quality of clad/frame and clad/fuel core, and the effect of these factors on products quality were comprehensively investigated and analyzed by means of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive X-ray spectrometry (EDX), image processing technique, etc.. The main results are as following: U-7.7%Si alloy contains two phases: primary U 3 Si 2 and small amount of USi (about 12%), free-uranium was not detected in fuel particles; the dogbone area is the key factor affecting fuel plate quality (1 ref., 16 figs., 4 tabs.)

Development method for measuring thickness of nuclei and coating of fuel plates

International Nuclear Information System (INIS)

Borges Junior, Reinaldo

2013-01-01

One of the most important components of a nuclear reactor is the Nuclear Fuel. Currently, the most advanced commercial fuel, whose applicability in Brazilian reactors has been developed by IPEN since 1985, is the silicide U 3 Si 2 . This is formed by fuel plates with nuclei dispersion (where the fissile material (U 3 Si 2 ) is homogeneously dispersed in a matrix of aluminum) coated aluminum. This fuel is produced in Brazil with developed technology, the result of the efforts made by the group of manufacturing nuclear fuel (CCN - Center of Nuclear Fuel) of IPEN. Considering the necessity of increasing the power of the IEA- R1 and Brazilian Multipurpose Reactor Building (RMB), for the production of radioisotopes - mainly for the area of medicine - there will be significant increase in the production of nuclear fuel at IPEN. Given this situation, if necessary, make the development of more modern and automated classification techniques. Aiming at this goal, this work developed a new computational method for measuring thickness of core and cladding of fuel plates, which are able to perform such measurements in less time and with more meaningful statistical data when compared with the current method of measurement. (author)

Flow channel shape optimum design for hydroformed metal bipolar plate in PEM fuel cell

Energy Technology Data Exchange (ETDEWEB)

Peng, Linfa; Lai, Xinmin; Liu, Dong' an; Hu, Peng [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China); Ni, Jun [Department of Mechanical Engineering and Applied Mechanics, University of Michigan, Ann Arbor, MI 48109 (United States)

2008-03-15

Bipolar plate is one of the most important and costliest components of polymer electrolyte membrane (PEM) fuel cells. Micro-hydroforming is a promising process to reduce the manufacturing cost of PEM fuel cell bipolar plates made of metal sheets. As for hydroformed bipolar plates, the main defect is the rupture because of the thinning of metal sheet during the forming process. The flow channel section decides whether high quality hydroformed bipolar plates can be successively achieved or not. Meanwhile, it is also the key factor that is related with the reaction efficiency of the fuel cell stacks. In order to obtain the optimum flow channel section design prior the experimental campaign, some key geometric dimensions (channel depth, channel width, rib width and transition radius) of flow channel section, which are related with both reaction efficiency and formability, are extracted and parameterized as the design variables. By design of experiments (DOE) methods and an adoptive simulated annealing (ASA) optimization method, an optimization model of flow channel section design for hydroformed metal bipolar plate is proposed. Optimization results show that the optimum dimension values for channel depth, channel width, rib width and transition radius are 0.5, 1.0, 1. 6 and 0.5 mm, respectively with the highest reaction efficiency (79%) and the acceptable formability (1.0). Consequently, their use would lead to improved fuel cell efficiency for low cost hydroformed metal bipolar plates. (author)

Fabrication of carbon-polymer composite bipolar plates for polymer electrolyte membrane fuel cells by compression moulding

International Nuclear Information System (INIS)

Raza, M.A.; Ahmed, R.; Saleem, A.; Din, R.U.

2009-01-01

Fuel cells are considered as one of the most important technologies to address the future energy and environmental pollution problems. These are the most promising power sources for road transportation and portable devices. A fuel cell is an electrochemical device that converts chemical energy into electrical energy. A fuel cell stack consists of bipolar plates and membrane electrode assemblies (MEA). The bipolar plate is by weight, volume and cost one of the most significant components of a fuel cell stack. Major functions of bipolar plates are to separate oxidant and fuel gas, provide flow channels, conduct electricity and provide heat transfer. Bipolar plates can be made from various materials including graphite, metals, carbon / carbon and carbon/ polymer composites. Materials for carbon-polymer composites are relatively inexpensive, less corrosive, strong and channels can be formed by means of a moulding process. Carbon-polymer composites are of two type i.e; thermosetting and thermoplastic. For thermosetting composite a bulk molding compound (BMC) was prepared by adding graphite, vinyl ester resin, methyl ethyl ketone peroxide and cobalt naphthalate. The BMC was thoroughly mixed, poured into a die mould of a bipolar plate with channels and hot pressed at a specific temperature and pressure. A bipolar plate was formed according to the die mould. Design of the mould is also discussed. Conducting polymers were also added to BMC to increase the conductivity of bipolar plates. Particle size of the graphite has also a significant effect on the conductivity of the bipolar plates. Thermoplastic composites were also prepared using polypropylene and graphite.

Estimate of fuel burnup spatial a multipurpose reactor in computer simulation

International Nuclear Information System (INIS)

Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

2015-01-01

In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

Corrosion of metal bipolar plates for PEM fuel cells: A review

Energy Technology Data Exchange (ETDEWEB)

Antunes, Renato A. [Engenharia de Materiais, Universidade Federal do ABC (UFABC), 09210-170 Santo Andre, SP (Brazil); Oliveira, Mara Cristina L.; Ett, Gerhard; Ett, Volkmar [Electrocell Ind. Com. Equip. Elet. LTDA, Centro de Inovacao, Empreendedorismo e Tecnologia (CIETEC), 05508-000 Sao Paulo, SP (Brazil)

2010-04-15

PEM fuel cells are of prime interest in transportation applications due to their relatively high efficiency and low pollutant emissions. Bipolar plates are the key components of these devices as they account for significant fractions of their weight and cost. Metallic materials have advantages over graphite-based ones because of their higher mechanical strength and better electrical conductivity. However, corrosion resistance is a major concern that remains to be solved as metals may develop oxide layers that increase electrical resistivity, thus lowering the fuel cell efficiency. This paper aims to present the main results found in recent literature about the corrosion performance of metallic bipolar plates. (author)

Design of metallic bipolar plates for PEM fuel cells.

Science.gov (United States)

2012-01-01

This project focused on the design and production of metallic bipolar plates for use in PEM fuel cells. Different metals were explored : and stainless steel was found out to be best suited to our purpose. Following the selection of metal, it was calc...

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

International Nuclear Information System (INIS)

Guigon, B.; Vacelet, H.; Dornbusch, D.

2000-01-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Design of the Fuel Element for the RRR Reactor (Australia)

International Nuclear Information System (INIS)

Estevez, E.A.; Markiewicz, M.E.; Gerding, R.

2003-01-01

The supply to the Replacement Research Reactor ( RRR ) to Australia represents a technological goal for our country, as much for the designers and manufacturers of this irradiation facility ( Invap SE ), as well for the responsibles of the fuel elements ( FE ) design and the suppliers of the first core ( CNEA ).In relation with the FE, although the conceptual design and fabrication technology of the FE are similar to the just developed and qualified by CNEA ( plane plates MTR fuel type ), the characteristics of this new reactor imposes most severe operation conditions on them than in previous supplies.In that sense, two distinguishing characteristics deserve to be shown: a) The magnitude of the hydrodynamics loads acting on the FE due to the coolant ascendent flow direction, and mainly, the very high flow velocities between the fuel plates ( aproximately five times higher than which presents in others Argentine FE actually in operation. b) The use of U3Si2 as fuel material.CNEA has started a programme to qualify this type of fuel.As result of these higher loads under irradiations and with the objective to maintain the high reliability level reached by our FE ( very low failure rates ), it was necessary to introduce FE mechanical-structural design modifications respect to the ECBE or standard design version, and to verify these changes through hydrodynamics tests on a 1:1 scale prototype.In this paper it is described the mechanical-structural FE design with special emphasis in the innovatives aspects incorporated.The design criteria established in function of the solicitations and limitating effects present under irradiation conditions.Also, a brief description of the proposed programme to verify and evaluate this design is presented, including analytical and numerical calculus of stresses acting on the fuel plates and others FE components, pressure loss hydrodynamics tests and endurance essays

Some tooling for manufacturing research reactor fuel plates

International Nuclear Information System (INIS)

Knight, R.W.

1999-01-01

This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment. (author)

Some Tooling for Manufacturing Research Reactor Fuel Plates

International Nuclear Information System (INIS)

Knight, R.W.

1999-01-01

This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment

End plate for e.g. solid oxide fuel cell stack, sets thermal expansion coefficient of material to predetermined value

DEFF Research Database (Denmark)

2011-01-01

.05-0.3 mm. USE - End plate for solid oxide fuel cell stack (claimed). Can also be used in polymer electrolyte fuel cell stack and direct methanol fuel cell stack. ADVANTAGE - The robustness of the end plate is improved. The structure of the end plate is simplified. The risk of delamination of the stack...

Combining different views of mammographic texture resemblance (MTR) marker of breast cancer risk

DEFF Research Database (Denmark)

Sun, S.; Karemore, Gopal; Chernoff, Konstantin

the subsequent 4 years whereas 245 cases had a diagnosis 2-4 years post mammography. We employed the MTR supervised texture learning framework to perform risk evaluation from a single mammography view. In the framework 20,000 pixels were sampled and classified by a kNN pixel classifier. A feature selection step......PURPOSE Mammographic density is a well established breast cancer risk factor. Texture analysis in terms of the Mammographoc Texture Resemblance (MTR) marker has recently shown to add to risk segregation. Hitherto only single view MTR analysis has been performed. Standard mammography examinations...

PcMtr, an aromatic and neutral aliphatic amino acid permease of Penicillium chrysogenum

NARCIS (Netherlands)

Trip, H; Evers, ME; Driessen, AJM

2004-01-01

The gene encoding an aromatic and neutral aliphatic amino acid permease of Penicillium chrysogenum was cloned, functionally expressed and characterized in Saccharomyces cerevisiae M4276. The permease, designated PcMtr, is structurally and functionally homologous to Mtr of Neurospora crassa, and

Mtr Extracellular Electron Transfer Pathways in Fe(III)-reducing or Fe(II)-oxidizing Bacteria: A Genomic Perspective

Energy Technology Data Exchange (ETDEWEB)

Shi, Liang; Rosso, Kevin M.; Zachara, John M.; Fredrickson, Jim K.

2012-12-01

Originally discovered in the dissimilatory metal-reducing bacterium Shewanella oneidensis MR-1 (MR-1), the Mtr (i.e., metal-reducing) pathway exists in all characterized strains of metal-reducing Shewanella. The protein components identified to date for the Mtr pathway of MR-1 include four multi-heme c-type cytochromes (c-Cyts), CymA, MtrA, MtrC and OmcA, and a porin-like, outer membrane protein MtrB. They are strategically positioned along the width of the MR-1 cell envelope to mediate electron transfer from the quinone/quinol pool in the inner-membrane to the Fe(III)-containing minerals external to the bacterial cells. A survey of microbial genomes revealed homologues of the Mtr pathway in other dissimilatory Fe(III)-reducing bacteria, including Aeromonas hydrophila, Ferrimonas balearica and Rhodoferax ferrireducens, and in the Fe(II)-oxidizing bacteria Dechloromonas aromatica RCB, Gallionella capsiferriformans ES-2 and Sideroxydans lithotrophicus ES-1. The widespread distribution of Mtr pathways in Fe(III)-reducing or Fe(II)-oxidizing bacteria emphasizes the importance of this type of extracellular electron transfer pathway in microbial redox transformation of Fe. Their distribution in these two different functional groups of bacteria also emphasizes the bi-directional nature of electron transfer reactions carried out by the Mtr pathways. The characteristics of the Mtr pathways may be shared by other pathways used by microorganisms for exchanging electrons with their extracellular environments.

Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

International Nuclear Information System (INIS)

Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

1995-01-01

A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

Postirradiation Examination Of U3O8-AL Plate Type Dispersion Fuel Element

International Nuclear Information System (INIS)

Nasution-Hasbullah; Sugondo; Amin, D.L.; Siti-Amini

1996-01-01

Postirradiation examination of plate type spent fuel element RIE-01 has been carried out in order to observer its physical changes and performance under irradiation in the reactor. The irradiation has been time more than two years with a declared burnup of 51.04 %. The examination included visual and dimensional measurement, measurement of burn-up distribution, wipe test and metallographic analysis. The results showed that all fuel plates retained their integrity. The colour changes were occurred on most of the plates significant suggesting that it was generated from the oxide layer formation. From gamma-scanning examination it could be deducted that the highest burn-up distribution of the plate was at position of 30 cm from the bottom. A more homogeneous distribution was found in the middle plate of the bundle. The increased plate thickness, as revealed by dimensional measurements as in agreement with the burn-up distribution pattern. Despite the changes observed in could be concluded that all changes occurred were still within the allowable limits and therefore it can recommended that an increase of the burn-up level above 51,04 % is still quite possible

Corrosion-resistant, electrically-conductive plate for use in a fuel cell stack

Science.gov (United States)

Carter, J David [Bolingbrook, IL; Mawdsley, Jennifer R [Woodridge, IL; Niyogi, Suhas [Woodridge, IL; Wang, Xiaoping [Naperville, IL; Cruse, Terry [Lisle, IL; Santos, Lilia [Lombard, IL

2010-04-20

A corrosion resistant, electrically-conductive, durable plate at least partially coated with an anchor coating and a corrosion resistant coating. The corrosion resistant coating made of at least a polymer and a plurality of corrosion resistant particles each having a surface area between about 1-20 m.sup.2/g and a diameter less than about 10 microns. Preferably, the plate is used as a bipolar plate in a proton exchange membrane (PEMFC) fuel cell stack.

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

Science.gov (United States)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

International Nuclear Information System (INIS)

Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

2011-01-01

A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

Corrosion and pyrophoricity of ZPPR fuel plates: Implications for basin storage

International Nuclear Information System (INIS)

Totemeier, T.C.; Hayes, S.L.; Pahl, R.G.; Crawford, D.C.

1997-01-01

This paper presents the results of recent experimentation and analysis of the pyrophoric behavior of corroded Zero Power Physics Reactor (ZPPR) HEU fuel plates and the implications of these results for the handling, drying, and passivation of uranium metal fuels stored in water basins. The ZPPR plates were originally clad in 1980; crevice corrosion of the uranium metal in a dry storage environment has occurred due to the use of porous cladding end plugs. The extensive corrosion has resulted in bulging and, in some cases, breaching of the cladding over a 15 year storage period. Processing of the plates has been initiated to recover the highly enriched uranium metal and remove the storage vulnerability identified with the corroded plates, which have been shown to contain significant quantities of the pyrophoric compound uranium hydride (UH 3 ). Experiments were undertaken to determine effective passivation techniques for the corrosion product; analysis and modeling was performed to determine whether heat generated by rapid hydride re-oxidation could ignite the underlying metal plates. The results of the initial passivation experiment showed that simple exposure of the hydride-containing corrosion product to an Ar-3 vol.% O 2 environment was insufficient to fully passivate the hydride--flare-up of the product occurred during subsequent vigorous handling in air. A second experiment demonstrated that corrosion product was fully stable following grinding of the product to a fine powder in the Ar-3 vol.% O 2 atmosphere. Numerical modeling of a corroded plate indicated that ignition of the plate due to the heat from hydride re-oxidation was likely if hydride fractions in the corrosion product exceeded 30%

Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

International Nuclear Information System (INIS)

Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

2013-01-01

The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

Thermal Characteristic Of AIMg2 Cladding And Fuel Plates Of U3Si2-Al With Various Uranium Loading

International Nuclear Information System (INIS)

Aslina, Br. G.; Suparjo; Aggraini, D.; Hasbullah, N.

1998-01-01

Thermal characteristic analyzed in this paper included linear expansion value, coefficient expansion, and enthalpy of cladding material fuel core and fuel plate of U 3 Si 2 -AI. Before analyzing, the fresh cladding of AIMg2 (without treatment) and the rolled AIMg2 were annealed at temperature of 425 o C for 1 hour, and the fuel plates of U 3 Si 2 -AI was prepared for various uranium loading of 0.9 - 3.6 - 4.2 - 4.8 and 5.2 g/cm 3 . Linear expansion nominal value and expansion coefficient were analyzed by using Dilatometer whereas enthalpy determination used Differential Thermal Analysis (DTA). The linear expansion and expansion coefficient analysis was performed to study the dimension cladding and of fuel plates during their stay in the reactor core, whereas determination of enthalpy was carried out to estimate the energy absorbed and released by fuel meat of U 3 Si 2 -AI to the cooling water through AlMg2 as a cladding. The result showed that the linear expansion and expansion coefficient of fresh AIMg2 cladding, rolled AIMg2 and fuel plates of U 3 Si 2 -AI are increased with the increase of temperature as well as the increase of uranium loading. The enthalpy measure showed that the enthalpy of fresh AIMg2 is smaller than that of rolled AIMg2 but melting temperature of fresh AIMg2 is greater than that of rolled AIMg2. The enthalpy of fuel plates and meat of U 3 Si 2 -AI is less than that of plates of U 3 Si 2 -AI. The enthalpy of fuel platers and meat of U 3 Si 2 -AI decrease with the increase of uranium loading. It is concluded that the fuel meat more reactive than fuel plates of U 3 Si 2 -AI

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)

Design of experiment study of the parameters that affect performance of three flow plate configurations of a proton exchange membrane fuel cell

International Nuclear Information System (INIS)

Carton, J.G.; Olabi, A.G.

2010-01-01

Low temperature hydrogen fuel cells are electrochemical devices which offer a promising alternative to traditional power sources. Fuel cells produce electricity with a reaction of the fuel (hydrogen) and air. Fuel cells have the advantage of being clean; only producing water and heat as by products. The efficiency of a fuel cell varies depending on the type; SOFC with CHP for example, can have a system efficiency of up to 65%. What the Authors present here is a comparison between three different configurations of flow plates of a proton exchange membrane fuel cell, the manufacturer's serpentine flow plate and two new configurations; the maze flow plate and the parallel flow plate. A study of the input parameters affecting output responses of voltage, current, power and efficiency of a fuel cell is performed through experimentation. The results were taken from direct readings of the fuel cell and from polarisation curves produced. This information was then analysed through a design of experiment to investigate the effects of the changing parameters on different configurations of the fuel cell's flow plates. The results indicate that, in relation to current and voltage response of the polarisation curve and the corresponding graphs produced from the DOE, the serpentine flow plate design is a much more effective design than the maze or parallel flow plate design. It was noted that the parallel flow plate performed reasonably well at higher pressures but over all statically the serpentine flow plate performed better.

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

International Nuclear Information System (INIS)

Sohn, Hee Dong

2012-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong

2012-02-15

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

Circular arc fuel plate stability experiments and analyses for the advanced neutron source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1995-08-01

The thin fuel plates planned for the Advanced Neutron Source are to be cooled by forcing heavy water at high velocity, 25 m/s, through thin cooling channels on each side of each plate. Because the potential for structural failure of the plates is a design concern, considerable effort has been expended in assessing this potential. As part of this effort, experimental flow tests and analyses to evaluate the structural response of circular arc plates have been conducted, and the results are given in this report

The velocity measurement by LDV at the simulated plate fuel assembly

International Nuclear Information System (INIS)

Tae Sung Ha

2001-01-01

For a more accurate safety analysis for McMaster Nuclear Reactor (MNR), local velocity measurements in a mock-up of the 18-plate fuel assembly are conducted over the range of M=2.0kg/s to 5.0kg/s (u=0.59m/s to 1.48m/s). To enable the measurement of the mass flow distribution through the channels by Laser Doppler Velocimeter(LDV), the curved fuel plate assembly is modified to flat fuel plates. The experimental result shows that the velocity profile is fairly symmetric for the 1st channel to the 17th subchannel at its center. The velocity in the peripheral area is slightly decreased while that directly above the circular pipe is correspondingly increased due to the effect of blockage by the exit endfitting. The mass flow rate fraction is fairly well distributed from the 1st to the 9th channels; at the outmost channels (1st and 3rd subchannels) the flow is approximately 95-97% of the average channel flow and at the central channels (4th and 8th subchannels) the flow is about 102-105% of the average channel mass flow rate. It is shown that the measured mass flow distribution is consistent with the results of the numerical calculation except 1st and 17th channels. (author)

What the difference to use LEU and HEU fuel elements separately or together in a research reactor

International Nuclear Information System (INIS)

Kaya, S.; Uestuen, G.

2005-01-01

Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates

Energy Technology Data Exchange (ETDEWEB)

Copeland, G.L.; Martin, M.M.

1980-12-01

A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

Shewanella putrefaciens mtrB encodes an outer membrane protein required for Fe(III) and Mn(IV) reduction.

Science.gov (United States)

Beliaev, A S; Saffarini, D A

1998-12-01

Iron and manganese oxides or oxyhydroxides are abundant transition metals, and in aquatic environments they serve as terminal electron acceptors for a large number of bacterial species. The molecular mechanisms of anaerobic metal reduction, however, are not understood. Shewanella putrefaciens is a facultative anaerobe that uses Fe(III) and Mn(IV) as terminal electron acceptors during anaerobic respiration. Transposon mutagenesis was used to generate mutants of S. putrefaciens, and one such mutant, SR-21, was analyzed in detail. Growth and enzyme assays indicated that the mutation in SR-21 resulted in loss of Fe(III) and Mn(IV) reduction but did not affect its ability to reduce other electron acceptors used by the wild type. This deficiency was due to Tn5 inactivation of an open reading frame (ORF) designated mtrB. mtrB encodes a protein of 679 amino acids and contains a signal sequence characteristic of secreted proteins. Analysis of membrane fractions of the mutant, SR-21, and wild-type cells indicated that MtrB is located on the outer membrane of S. putrefaciens. A 5.2-kb DNA fragment that contains mtrB was isolated and completely sequenced. A second ORF, designated mtrA, was found directly upstream of mtrB. The two ORFs appear to be arranged in an operon. mtrA encodes a putative 10-heme c-type cytochrome of 333 amino acids. The N-terminal sequence of MtrA contains a potential signal sequence for secretion across the cell membrane. The amino acid sequence of MtrA exhibited 34% identity to NrfB from Escherichia coli, which is involved in formate-dependent nitrite reduction. To our knowledge, this is the first report of genes encoding proteins involved in metal reduction.

Recent status and future aspect of plate type fuel element technology with high uranium density at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.F.; Hassel, H.-W.

1983-01-01

According to the present state of development full size test fuel elements with UAl x , U 3 O 8 , and U 3 Si 2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAI x , 2.9 g/cc for U 3 O 8 , and 4.76 g/cc for U 3 Si 2 . The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1 x , 60% with U 3 O 8 , 39% with U 3 Si 2 ; no defects could be detected. For an economical fuel element production with reduced 235-U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl x and U 3 O 8 fuel were successfully irradiated in the ORR up to a burnup of 75 %. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAI x , and 3.14 g/cc for U 3 O 8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)

The Conserved Actinobacterial Two-Component System MtrAB Coordinates Chloramphenicol Production with Sporulation in Streptomyces venezuelae NRRL B-65442

Directory of Open Access Journals (Sweden)

Nicolle F. Som

2017-06-01

Full Text Available Streptomyces bacteria make numerous secondary metabolites, including half of all known antibiotics. Production of antibiotics is usually coordinated with the onset of sporulation but the cross regulation of these processes is not fully understood. This is important because most Streptomyces antibiotics are produced at low levels or not at all under laboratory conditions and this makes large scale production of these compounds very challenging. Here, we characterize the highly conserved actinobacterial two-component system MtrAB in the model organism Streptomyces venezuelae and provide evidence that it coordinates production of the antibiotic chloramphenicol with sporulation. MtrAB are known to coordinate DNA replication and cell division in Mycobacterium tuberculosis where TB-MtrA is essential for viability but MtrB is dispensable. We deleted mtrB in S. venezuelae and this resulted in a global shift in the metabolome, including constitutive, higher-level production of chloramphenicol. We found that chloramphenicol is detectable in the wild-type strain, but only at very low levels and only after it has sporulated. ChIP-seq showed that MtrA binds upstream of DNA replication and cell division genes and genes required for chloramphenicol production. dnaA, dnaN, oriC, and wblE (whiB1 are DNA binding targets for MtrA in both M. tuberculosis and S. venezuelae. Intriguingly, over-expression of TB-MtrA and gain of function TB- and Sv-MtrA proteins in S. venezuelae also switched on higher-level production of chloramphenicol. Given the conservation of MtrAB, these constructs might be useful tools for manipulating antibiotic production in other filamentous actinomycetes.

Tensile mechanical properties of U3Si2-Al fuel plate

International Nuclear Information System (INIS)

Xu Yong; Hu Huawei; Zhuang Hongquan; Wang Xishu

2003-01-01

The fuel plate made of fuel meat, with the U 3 Si 2 -Al dispersion fuel center, and 6061 Al alloy cladding, is a new kind of fuel used in research reactors. The mechanical property data of the fuel meat is the basic data in the design of fuel group, but the mechanical property of this fuel meat has not been studied all over the world till now. In this paper, the mechanical properties of U 3 Si 2 -Al fuel meats of different sizes used in research reactors are investigated and analyzed, and at the same time the carrying capacity of tensile in different directions are also compared. In order to get more knowledge about the mechanical properties of the fuel meat, the tensile experiment has been carried out repeatedly. Considering the lower ratio of elongation and the brittleness, the microscope has been used to examine the zone of fracture after tensile test. (authors)

Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

1991-08-01

According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

2016-10-15

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

Study on the Applicability of Electron Beam Welding Methods to Assembly a Fuel Compact and Al Cover Plate of Research Reactor Plate Type Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Hae In; Lee, Yoon Sang; Lee, Don Dae; Jeong, Yong Jin; Kwon, Sun Chil; Kim, Soo Sung; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Among the research reactor plate type fuel fabrication processes, there is an assembly process between fuel meat compact and Al cover plates using a welding method prior to rolling process. The assembly process is such as the Al frame and Al cover plate should be welded properly as shown in Fig. 1. For welding, TIG(Tungsten Inert Gas) welding methods has been used conventionally, but in this study an electron beam welding(EB welding) technique which uses the electron beam of a high velocity for joining two materials is introduced to the assembly. The work pieces are melted as the kinetic energy of the electron beam is transformed into heat to join the two parts of the weld. The welding is often done in the conditions in a vacuum to prevent dispersion of the electron beam. The electron beam welding process has many ad-vantages such as contamination of the welds could be prevented, the penetration of the weld is deep, and also the strain of the welding area is less than other methods. In this study, to find optimal condition of the EB welding process, a welding speed, a beam current and an acceleration voltage were changed. To analyzing the welding results, the shape of the beads and defects of welding area was used. The width and depth of the beads were measured as well

Study on the Applicability of Electron Beam Welding Methods to Assembly a Fuel Compact and Al Cover Plate of Research Reactor Plate Type Fuel

International Nuclear Information System (INIS)

Lee, Hae In; Lee, Yoon Sang; Lee, Don Dae; Jeong, Yong Jin; Kwon, Sun Chil; Kim, Soo Sung; Park, Jong Man

2012-01-01

Among the research reactor plate type fuel fabrication processes, there is an assembly process between fuel meat compact and Al cover plates using a welding method prior to rolling process. The assembly process is such as the Al frame and Al cover plate should be welded properly as shown in Fig. 1. For welding, TIG(Tungsten Inert Gas) welding methods has been used conventionally, but in this study an electron beam welding(EB welding) technique which uses the electron beam of a high velocity for joining two materials is introduced to the assembly. The work pieces are melted as the kinetic energy of the electron beam is transformed into heat to join the two parts of the weld. The welding is often done in the conditions in a vacuum to prevent dispersion of the electron beam. The electron beam welding process has many ad-vantages such as contamination of the welds could be prevented, the penetration of the weld is deep, and also the strain of the welding area is less than other methods. In this study, to find optimal condition of the EB welding process, a welding speed, a beam current and an acceleration voltage were changed. To analyzing the welding results, the shape of the beads and defects of welding area was used. The width and depth of the beads were measured as well

The improvement of technology for high-uranium-density Al-base dispersion fuel plates

International Nuclear Information System (INIS)

Shouhui, Dai; Rongxian, Sun; Hejian, Mao; Baosheng, Zhao; Changgen, Yin

1987-01-01

An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

Fabrication of high-uranium-loaded U{sub 3}O{sub 8}-Al developmental fuel plates

Energy Technology Data Exchange (ETDEWEB)

Copeland, G L; Martin, M M [Oak Ridge National Laboratory, TN (United States)

1983-08-01

A common plate-type fuel for Research and Test Reactors (RERTR) is U{sub 3}0{sub 8} dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the {sup 235}U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for non-peaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service. We fabricated developmental fuel plates with cores containing from 60 to 100 wt U{sub 3}0{sub 8} in aluminum encapsulated in 6061 aluminum alloy and evaluated them for aspects of fabricability, nondestructive testing, and expected performance. We recommend 75 wt U{sub 3}0{sub 8}-Al 3.1 Mg U/m{sup 3}) as the highest loading in the initial irradiation test. This upper limit is based on a qualitative assessment of the mechanical integrity of the core made by using current fabrication techniques and materials. As the oxide loading is increased beyond this point, planar areas and extensive stringers of oxide and voids develop, which leave little strength in the thickness direction. Fuel plates may then blister over these areas as fission gases collect during irradiation. Current size plates are easily fabricable to the 75 wt % U{sub 3}0{sub 8}-Al core loading by current fabrication techniques. Dogboning is a potential problem at this loading for some applications; however, this can be easily solved by using tapered compact ends. Current nondestructive radiography and transmission x-ray scanning are applicable to the highly loaded plates. Ultrasonic testing for non-bonds is marginal because of the abrupt change in conductance at the cladding-core interface. Plate thickness can be increased if desired; we fabricated 75 wt % plates with

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T [ORNL

2014-05-01

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

Uranium density reduction on fuel element side plates assessment

International Nuclear Information System (INIS)

Rios, Ilka A.; Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E.

2011-01-01

During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

Uranium density reduction on fuel element side plates assessment

Energy Technology Data Exchange (ETDEWEB)

Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

Corrosion resistance characteristics of stamped and hydroformed proton exchange membrane fuel cell metallic bipolar plates

Energy Technology Data Exchange (ETDEWEB)

Dundar, F. [NSF I/UCRC Center for Precision Forming (CPF), Virginia Commonwealth University, Richmond, VA (United States); Department of Materials Science and Engineering, Gebze Institute of Technology (Turkey); Dur, Ender; Koc, M. [NSF I/UCRC Center for Precision Forming (CPF), Virginia Commonwealth University, Richmond, VA (United States); Mahabunphachai, S. [NSF I/UCRC Center for Precision Forming (CPF), Virginia Commonwealth University, Richmond, VA (United States); National Metal and Materials Technology Center (MTEC), Pathumthani (Thailand)

2010-06-01

Metallic bipolar plates have several advantages over bipolar plates made from graphite and composites due to their high conductivity, low material and production costs. Moreover, thin bipolar plates are possible with metallic alloys, and hence low fuel cell stack volume and mass are. Among existing fabrication methods for metallic bipolar plates, stamping and hydroforming are seen as prominent approaches for mass production scales. In this study, the effects of important process parameters of these manufacturing processes on the corrosion resistance of metallic bipolar plates made of SS304 were investigated. Specifically, the effects of punch speed, pressure rate, stamping force and hydroforming pressure were studied as they were considered to inevitably affect the bipolar plate micro-channel dimensions, surface topography, and hence the corrosion resistance. Corrosion resistance under real fuel cell conditions was examined using both potentiodynamic and potentiostatic experiments. The majority of the results exhibited a reduction in the corrosion resistance for both stamped and hydroformed plates when compared with non-deformed blank plates of SS304. In addition, it was observed that there exist an optimal process window for punch speed in stamping and the pressure rate in hydroforming to achieve improved corrosion resistance at a faster production rate. (author)

Post-irradiation studies of test plates for low enriched fuel elements for research reactors

International Nuclear Information System (INIS)

Groos, E.; Buecker, H.J.; Derz, H.; Schroeder, R.

1988-07-01

In developing new fuels for research reactor elements that allow the use of low enriched uranium (LEU) 3 Si 2 , U 3 Si 1.5 , U 3 Si 1.3 and U 3 Si. Even up to high burnup rates (80% fifa) U 3 Si 2 was proved to be a reliable fuel that according to the test results achieved to date complies with all necessary requirements above all with respect to dimensional stability. U 3 Si showed significant changes of the fuel microstructure associated with considerably higher fuel swelling, that will probably exclude its use in research reactor operation. The irradiation of U 3 Si 1.3 and U 3 Si 1.5 plates had to be terminated untimely. Up to a burnup of 40% fifa these plates behaved quite well. An extrapolation to higher burnup rates, however only seems to be possible with reservations. (orig./HP) [de

Study on characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates

International Nuclear Information System (INIS)

Liu Lijian; Yin Changgeng; Chen Jiangang; Sun Changlong; Liu Yunming

2014-01-01

In this paper, we analyzed the characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates. The results show that the interaction layers (IL) are with irregular morphology and uneven thickness, and are mainly formed in the internal micro cracks of the dispersion fuel particles or at the interface between the particles and the substrates. The diffusion mechanism of U-Mo/Al-Si is the vacancy diffusion, Al and Si are migrating elements, and the diffusion reaction is that Al and Si diffuse to U-Mo alloy. Inside the interaction layers, the Al content keeps constant basically, but the Si content gradually increases with the substrate-fuel direction, and the maximum content of Si appears interaction layers near the U-Mo side. Adding about 5 wt% Si into Al matrix can restrain the diffusion reaction, and improve the performance of dispersion fuel plates finally. (authors)

Effect of pervaporation plate thickness on the rate of methanol evaporation in a passive vapor-feed direct methanol fuel cell

Science.gov (United States)

Fauzi, N. F. I.; Hasran, U. A.; Kamarudin, S. K.

2015-09-01

In a passive vapor-feed direct methanol fuel cell (DMFC), methanol vapor is typically obtained using a pervaporation plate in a process by which liquid methanol contained in the fuel reservoir undergoes a phase change to vapor in the anodic vapor chamber. This work investigates the effect of pervaporation plate thickness on the rate of methanol evaporation using a three-dimensional simulation model developed by varying the plate thickness. A. The rate of methanol evaporation was measured using Darcy's law. The rate of methanol evaporation was found to be inversely proportional to the plate thickness, where the decrease in thickness inevitably lowers the resistance along the plate and consequently increases the methanol transport through the plate. This shows that the plate thickness has a significant influence on the rate of methanol evaporation and thereby plays an important role in improving the performance of the passive vapor-feed direct methanol fuel cell.

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

Energy Technology Data Exchange (ETDEWEB)

Leenaers, A., E-mail: aleenaer@sckcen.be [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van den Berghe, S.; Koonen, E.; Kuzminov, V. [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, 9000 Ghent (Belgium)

2015-03-15

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK• CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% {sup 235}U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

Science.gov (United States)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

2015-03-01

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

International Nuclear Information System (INIS)

Eko Priyono; Saminto

2016-01-01

Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

Use of gamma spectrometry for studying fuel plates

International Nuclear Information System (INIS)

Carteret, Y.; Schley, R.; Simonet, G.

1979-01-01

The programme of experimental irradiation performed at the CEA on the CARAMEL plate fuel was followed by gamma spectrometry, jointly with other techniques. The qualitative study of the distribution of fission products constitutes a source of information on the behavior of the fuel (temperature and structure) and enables its utilization limits to be predicted. The quantitative determination of short and long half life fission products makes it possible to calculate the specific power and specific burn-up. Carried out periodically, it is a means of checking the values obtained by the continuous measurement of cladding temperature, directly linked to the specific burn-up. At the end of irradiation, the results are compared against those achieved by neodymium analysis. The study of the change in gadolinium, a burnable poison, is an application of this technique [fr

Effect of pervaporation plate thickness on the rate of methanol evaporation in a passive vapor-feed direct methanol fuel cell

International Nuclear Information System (INIS)

Fauzi, N F I; Hasran, U A; Kamarudin, S K

2015-01-01

In a passive vapor-feed direct methanol fuel cell (DMFC), methanol vapor is typically obtained using a pervaporation plate in a process by which liquid methanol contained in the fuel reservoir undergoes a phase change to vapor in the anodic vapor chamber. This work investigates the effect of pervaporation plate thickness on the rate of methanol evaporation using a three-dimensional simulation model developed by varying the plate thickness. A. The rate of methanol evaporation was measured using Darcy's law. The rate of methanol evaporation was found to be inversely proportional to the plate thickness, where the decrease in thickness inevitably lowers the resistance along the plate and consequently increases the methanol transport through the plate. This shows that the plate thickness has a significant influence on the rate of methanol evaporation and thereby plays an important role in improving the performance of the passive vapor-feed direct methanol fuel cell. (paper)

Full-sized plates irradiation with high UMo fuel loading. Final results of IRIS 1 experiment

International Nuclear Information System (INIS)

Huet, F.; Marelle, V.; Noirot, J.; Sacristan, P.; Lemoine, P.

2003-01-01

As a part of the French UMo Group qualification program, IRIS 1 experiment contained full-sized plates with high uranium loading in the meat of 8 g.cm -3 . The fuel particles consisted of 7 and 9 wt% Mo-uranium alloys ground powders. The plate were irradiated at OSIRIS reactor in IRIS device up to 67.5% peak burnup within the range of 136 W.cm - '2 for the heat flux and 72 deg. C for the cladding temperature. After each reactor cycle the plates thickness were measured. The results show no swelling behaviour differences versus burnup between UMo7 and UMo9 plates. The maximum plate swelling for peak burnup location remains lower than 6%. The wide set of PIE has shown that, within the studied irradiation conditions, the interaction product have a global formulation of '(U-Mo)Al -7 ' and that there is no aluminium dissolution in UMo particles. IRIS1 experiment, as the first step of the UMo fuel qualification for research reactor, has established the good behaviour of UMo7 and UMo9 high uranium loading full-sized plate within the tested conditions. (author)

Structure and Function of Neisseria gonorrhoeae MtrF Illuminates a Class of Antimetabolite Efflux Pumps

Directory of Open Access Journals (Sweden)

Chih-Chia Su

2015-04-01

Full Text Available Neisseria gonorrhoeae is an obligate human pathogen and the causative agent of the sexually transmitted disease gonorrhea. The control of this disease has been compromised by the increasing proportion of infections due to antibiotic-resistant strains, which are growing at an alarming rate. N. gonorrhoeae MtrF is an integral membrane protein that belongs to the AbgT family of transporters for which no structural information is available. Here, we describe the crystal structure of MtrF, revealing a dimeric molecule with architecture distinct from all other families of transporters. MtrF is a bowl-shaped dimer with a solvent-filled basin extending from the cytoplasm to halfway across the membrane bilayer. Each subunit of the transporter contains nine transmembrane helices and two hairpins, posing a plausible pathway for substrate transport. A combination of the crystal structure and biochemical functional assays suggests that MtrF is an antibiotic efflux pump mediating bacterial resistance to sulfonamide antimetabolite drugs.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

Science.gov (United States)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

1992-01-01

The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

Utilization of radiographic and ultrasonic testing for an evaluation of plate type fuel elements during manufacturing stages

International Nuclear Information System (INIS)

Brito, Mucio Jose Drummond de; Silva Junior, Silverio Ferreira da; Messias, Jose Marcos; Braga, Daniel Martins; Paula, Joao Bosco de

2005-01-01

Structural discontinuities can be introduced in the plate type fuel elements during the manufacturing stages due to mechanical processing conditions. The use of nondestructive testing methods to monitoring the fuel elements during the manufacturing stages presents a significant importance, contributing for manufacturing process improvement and cost reducing. This paper describes a procedure to be used detection and evaluation of structural discontinuities in plate type fuel elements during the manufacturing stages using the ultrasonic testing method and the radiographic testing method. The main results obtained are presented and discussed. (author)

PLACA/DPLACA: a code to simulate the behavior of a monolithic/dispersed plate type fuel

International Nuclear Information System (INIS)

Denis, Alicia; Soba, Alejandro

2005-01-01

The PLACA code was originally built to simulate monolithic plate fuels contained in a metallic cladding, with a gap in between. The international program of high density fuels was recently oriented to the development of a plate-type fuel of a uranium rich alloy with a molybdenum content between 6 to 10 w %, without gap and with a Zircaloy cladding. To give account of these fuels, the DPLACA code was elaborated as a modification of the original code. The extension of the calculation tool to disperse fuels involves a detailed study of the properties and models (still in progress). Of special interest is the material formed by U Mo particles dispersed in an Al matrix. This material has appeared as a candidate fuel for high flux research reactors. However, the interaction layer that grows around the particles has a deleterious effect on the material performance in operation conditions and may represent a limit for its applicability. A number of recent experiments carried out on this material provide abundant information that allows testing of the numerical models. (author)

Programming the quorum sensing-based AND gate in Shewanella oneidensis for logic gated-microbial fuel cells.

Science.gov (United States)

Hu, Yidan; Yang, Yun; Katz, Evgeny; Song, Hao

2015-03-11

An AND logic gate based on a synthetic quorum-sensing (QS) module was constructed in a Shewanella oneidensis MR-1 mtrA knockout mutant. The presence of two input signals activated the expression of a periplasmic decaheme cytochrome MtrA to regenerate the extracellular electron transfer conduit, enabling the construction of AND-gated microbial fuel cells.

3D COMSOL Simulations for Thermal Deflection of HFIR Fuel Plate in the "Cheverton-Kelley" Experiments

Energy Technology Data Exchange (ETDEWEB)

Jain, Prashant K [ORNL; Freels, James D [ORNL; Cook, David Howard [ORNL

2012-08-01

Three dimensional simulation capabilities are currently being developed at Oak Ridge National Laboratory using COMSOL Multiphysics, a finite element modeling software, to investigate thermal expansion of High Flux Isotope Reactor (HFIR) s low enriched uranium fuel plates. To validate simulations, 3D models have also been developed for the experimental setup used by Cheverton and Kelley in 1968 to investigate the buckling and thermal deflections of HFIR s highly enriched uranium fuel plates. Results for several simulations are presented in this report, and comparisons with the experimental data are provided when data are available. A close agreement between the simulation results and experimental findings demonstrates that the COMSOL simulations are able to capture the thermal expansion physics accurately and that COMSOL could be deployed as a predictive tool for more advanced computations at realistic HFIR conditions to study temperature-induced fuel plate deflection behavior.

Research reactor fuel transport in the U.K

Energy Technology Data Exchange (ETDEWEB)

Panter, R [U.K. Atomic Energy Authority, Harwell (United Kingdom)

1983-09-01

This paper describes the containers currently used for transport of fresh or spent fuel elements for Research and Materials Test Reactors in the U.K., their status, operating procedures and some of the practical difficulties. In the U.K., MTR fuel cycle work is almost entirely the responsibility of the U.K. Atomic Energy Authority.

Micro direct methanol fuel cell with perforated silicon-plate integrated ionomer membrane

DEFF Research Database (Denmark)

Larsen, Jackie Vincent; Dalslet, Bjarke Thomas; Johansson, Anne-Charlotte Elisabeth Birgitta

2014-01-01

This article describes the fabrication and characterization of a silicon based micro direct methanol fuel cell using a Nafion ionomer membrane integrated into a perforated silicon plate. The focus of this work is to provide a platform for micro- and nanostructuring of a combined current collector...... at a perforation ratio of 40.3%. The presented fuel cells also show a high volumetric peak power density of 2 mW cm−3 in light of the small system volume of 480 μL, while being fully self contained and passively feed....... and catalytic electrode. AC impedance spectroscopy is utilized alongside IV characterization to determine the influence of the plate perforation geometries on the cell performance. It is found that higher ratios of perforation increases peak power density, with the highest achieved being 2.5 mW cm−2...

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Craft, Aaron E.; Papaioannou, Glen C.; Chichester, David L.; Williams, Walter J.

2017-02-01

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutron radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.

Loading procedures for shipment of irradiated fuel

Energy Technology Data Exchange (ETDEWEB)

Bates, E F; Feltz, D E; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

1974-07-01

The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

Loading procedures for shipment of irradiated fuel

International Nuclear Information System (INIS)

Bates, E.F.; Feltz, D.E.; Sandel, P.S.; Schoenbucher, B.

1974-01-01

The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

Energy Technology Data Exchange (ETDEWEB)

Andrzejewski, Claudio de Sa

2005-07-01

Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

Stress Linearization and Strength Evaluation of the BEP's Flow Plates for a Dual Cooled Fuel Assembly

International Nuclear Information System (INIS)

Kim, Jae Yong; Yoon, Kyung Ho; Kang, Heung Seok; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

2009-01-01

A fuel assembly is composed of 5 major components, such as a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT) and fuel rods (FRs). There are no ASME criteria about all components except for a TEP/BEP. The TEP/BEP should satisfy stress intensity limits in case of condition A and B of ASME, Section III, Division 1 . Subsection NB. In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape to provide flow holes to direct the heated coolant into/out of the fuel assembly but structural intensity of these plates within a 22.241 kN axial loading should satisfy Tresca stress limits in ASME code. In this paper, stress linearization procedure and strength evaluation of a newly designed BEP for the dual cooled fuel assembly are described

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

International Nuclear Information System (INIS)

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

1996-05-01

As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

Polymer electrolyte membrane fuel cell (PEMFC) flow field plate: design, materials and characterisation

Energy Technology Data Exchange (ETDEWEB)

Hamilton, P.J.; Pollet, B.G. [PEM Fuel Cell Research Group, School of Chemical Engineering, University of Birmingham, Edgbaston, B15 2TT (United Kingdom)

2010-08-15

This review describes some recent developments in the area of flow field plates (FFPs) for proton exchange membrane fuel cells (PEMFCs). The function, parameters and design of FFPs in PEM fuel cells are outlined and considered in light of their performance. FFP materials and manufacturing methods are discussed and current in situ and ex situ characterisation techniques are described. (Abstract Copyright [2010], Wiley Periodicals, Inc.)

Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

Energy Technology Data Exchange (ETDEWEB)

Perez, E.; Yao, B. [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States); Keiser, D.D. [Nuclear Fuels and Materials Division, Idaho National Laboratory, Scoville, ID 83415 (United States); Sohn, Y.H., E-mail: ysohn@mail.ucf.ed [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States)

2010-07-01

For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr{sub 2}, {gamma}-UZr, Zr solid-solution and Mo{sub 2}Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si){sub 2}Zr, (Al, Si)Zr{sub 3} (Al, Si){sub 3}Zr, and AlSi{sub 4}Zr{sub 5}. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

Science.gov (United States)

Perez, E.; Yao, B.; Keiser, D. D., Jr.; Sohn, Y. H.

2010-07-01

For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

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.

Bipolar plate materials in molten carbonate fuel cells. Final CRADA report.

Energy Technology Data Exchange (ETDEWEB)

Krumpelt, M.

2004-06-01

Advantages of implementation of power plants based on electrochemical reactions are successfully demonstrated in the USA and Japan. One of the msot promising types of fuel cells (FC) is a type of high temperature fuel cells. At present, thanks to the efforts of the leading countries that develop fuel cell technologies power plants on the basis of molten carbonate fuel cells (MCFC) and solid oxide fuel cells (SOFC) are really close to commercialization. One of the problems that are to be solved for practical implementation of MCFC and SOFC is a problem of corrosion of metal components of stacks that are assembled of a number of fuel cells. One of the major components of MCFC and SOFC stacks is a bipolar separator plate (BSP) that performs several functions - it is separation of reactant gas flows sealing of the joints between fuel cells, and current collection from the surface of electrodes. The goal of Task 1 of the project is to develop new cost-effective nickel coatings for the Russian 20X23H18 steel for an MCFC bipolar separator plate using technological processes usually implemented to apply corrosion stable coatings onto the metal parts for products in the defense. There was planned the research on production of nickel coatings using different methods, first of all the galvanic one and the explosion cladding one. As a result of the works, 0.4 x 712 x 1296 mm plates coated with nickel on one side were to be made and passed to ANL. A line of 4 galvanic baths 600 liters was to be built for the galvanic coating applications. The goal of Task 2 of the project is the development of a new material of an MCFC bipolar separator plate with an upgraded corrosion stability, and development of a technology to produce cold roll sheets of this material the sizes of which will be 0.8 x 712x 1296 mm. As a result of these works, a pilot batch of the rolled material in sheets 0.8 x 712 x 1296 mm in size is to be made (in accordance with the norms and standards of the Russian

Effect of the fabrication process on fatigue performance of U3Si2 fuel plate with sandwich structure

International Nuclear Information System (INIS)

Wang Xishu; Li Shuangshou; Wang Qingyuan; Xu Yong

2005-01-01

U 3 Si 2 -Al fuel plate is one of the dispersion fuel structure materials recently developed and widely used in research reactors. The mechanical properties of this structural material, especially the fatigue performance, are strongly dependent on its fabrication process. To investigate the effects of these processing technologies, the fatigue tests for the different specimens were carried out. The S-N curves indicate that the fabrication processing technologies of U 3 Si 2 fuel plate, such as the addition of U 3 Si 2 particles into aluminum powder to form the fuel meat, holding and rolling the processes of meat and cladding of 6061-Al alloy, plays an important role in improving the mechanical properties and fatigue performance of this fuel plate. In addition, some factors that influence the crack initiation and propagation are summarized based on the fatigue images that are in situ observations with SEM. The critical criterion for fatigue damage is proposed based on the fatigue data of the structural material, which were obtained at the different conditions

Detection of mutations in mtrR gene in quinolone resistant strains of N.gonorrhoeae isolated from India

Directory of Open Access Journals (Sweden)

S V Kulkarni

2015-01-01

Full Text Available Background and Objectives: Emergence of multi-drug resistant Neisseria gonorrhoeae resulting from new genetic mutation is a serious threat in controlling gonorrhea. This study was undertaken to identify and characterise mutations in the mtrR genes in N.gonorrhoeae isolates resistant to six different antibiotics in the quinolone group. Materials and Methods: The Minimum inhibitory concentrations (MIC of five quinolones for 64 N.gonorrhoeae isolates isolated during Jan 2007-Jun 2009 were determined by E-test method. Mutations in MtrR loci were examined by deoxyribonucleic acid (DNA sequencing. Results: The proportion of N.gonorrhoeae strains resistant to anti-microbials was 98.4% for norfloxacin and ofloxacin, 96.8% for enoxacin and ciprofloxacin, 95.3% for lomefloxacin. Thirty-one (48.4% strains showed mutation (single/multiple in mtrR gene. Ten different mutations were observed and Gly-45 → Asp, Tyr-105 → His being the most common observed mutation. Conclusion: This is the first report from India on quinolone resistance mutations in MtrRCDE efflux system in N.gonorrhoeae. In conclusion, the high level of resistance to quinolone and single or multiple mutations in mtrR gene could limit the drug choices for gonorrhoea.

Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

Energy Technology Data Exchange (ETDEWEB)

Rakesh, R., E-mail: rakesh.rad87@gmail.com [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Kohli, D. [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Sinha, V.P.; Prasad, G.J. [Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Samajdar, I. [Department of Metallurgical Engineering and Materials Science, IIT Bombay, Powai, Mumbai 400076 (India)

2014-02-01

Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum–aluminum (case A) and aluminum–aluminum + yttria (Y{sub 2}O{sub 3}) dispersion (case B). Case B approximated aluminum–uranium silicide (U{sub 3}Si{sub 2}) ‘fuel-meat’ in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in ‘out-of-plane’ residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

analysis of reactivity accidents in MTR for various protection system parameters and core condition

International Nuclear Information System (INIS)

Mohamed, F.M.

2011-01-01

Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

Application of the beta particles backscattering technique for determining the thickness of the cladding in nuclear fuels plate

International Nuclear Information System (INIS)

Koshimizu, S.; Ferreira, P.I.; Lima, L.F.C.P. de; Vieira, J.M.; Perez, H.E.B.

1984-01-01

A prototype of an instalation to measure thickness of cladding and core of nuclear fuels plate using the beta particles backscattering technique is constructed. The method and calibration system is described. The thickness measurements of the cladding and core were done in a natural uranium fuel plate developed at IPEN. The reliability of the method is confirmed by the metalographic measures analysis. (E.G.) [pt

Does Magnetization Transfer Ratio (MTR) contribute to the diagnosis and differential diagnosis of the dementias?

International Nuclear Information System (INIS)

Hentschel, F.; Kreis, M.; Damian, M.; Krumm, B.

2004-01-01

Purpose: The magnetization transfer ratio (MTR) is a MR-based neuroimaging procedure aiming at the quantification of the structural integrity of brain tissue. Its contribution to the differential diagnosis of dementias was examined and discussed in relation to the pathogenesis of age-related dementias. Materials and Methods: Sixty-one patients from a memory clinic were diagnosed by general physical and neuropsychiatric examination, and underwent neuropsychologic testing and neuroimaging using MRI. Their clinical diagnoses were based on standard operational research criteria. Additionally, the MTR in 10 defined regions of interest (ROI) was determined. This investigation was performed using a T1-weighted SE sequence. Average MTR values were determined in the individual ROI and their combinations and correlated with the age gender, cognitive impairment and clinical diagnosis. Sensitivity, specificity, positive and negative predictive value were determined, as well as the rate of correct classifications. Results: For cognitive healthy subjects, the MRT values correlate only mildly, though significantly, with age in the hippocampus and with gender in the dorsal corpus callosum. In contrast, the MTR in the frontal white matter correlates strongly and highly significantly with cognitive impairment in patients with dementia. The differential diagnostic assignment of Alzheimer's disease versus vascular dementia by MTR provides a correct classification of approximately 50% to 70%. PPV for no dementia vs. vascular dementia or the NPV for vascular vs. Alzheimer's disease are considerably higher exceeding 80%. For no dementia vs. Alzheimer's disease, the NPV was over 90%. (orig.)

Implementation of the non-destructive ultrasound testing by immersion through the transmission technique, applied to the quality control of nuclear fuel plates

International Nuclear Information System (INIS)

Medina Jofre, David Christian

2014-01-01

Within the framework of global development, which seeks to reduce the enrichment of U 235 in nuclear fuels for research reactors, the Fuel Elements Plant (PEC) of the Chilean Nuclear Energy Commission (CCHEN) has worked with the Idaho National Laboratory (INL-USA), for the fabrication of high density fuel plates based on the dispersion of Uranium-Molybdenum alloy powders (UMo), which are subjected to inspections and tests to qualify as a compliant product for use in nuclear research reactors. It is in this matter where the Non Destructive Test (NDT) of immersion ultrasound used in both facilities differs in its acceptance criteria, when is used different testing techniques; On the one hand, the PEC uses the pulse-echo technique, while the INL uses the transmission technique. Therefore, the present work is focused on the implementation of the ultrasound by immersion using the transmission technique. During the development of the work, the physical and virtual configuration of the ultrasound equipment was possible and elaborate an operation procedure, which allows to inspect through this technique, a series of fuel plates based on UMo and U 3 Si 2 powders, with different characteristics. The results allow to characterize the signals obtained in fuel plates according to the nuclear fuel material used. There is an inverse relationship between the uranium load per unit volume (uranium density, gU/cm 3 ) used in the fuel plate and the transmittance of the ultrasonic beam through the areas where there is nuclear fuel material (meat); the effect produced by a dispersed combustible material is observed and it is possible to identify discontinuities that may be present in the fuel plate. Finally, an inspection technical instruction for U 3 Si 2 fuel plates is elaborated, where acceptance and rejection criteria are defined

Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

International Nuclear Information System (INIS)

Setiyanto; Tukiran Surbakti

2016-01-01

In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

OpenAIRE

Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

2016-01-01

Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

Irradiation experience of IPEN fuel at IEA-R1 research reactor

International Nuclear Information System (INIS)

Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

Influence of the silicon content on the core corrosion properties of dispersion type fuel plates

International Nuclear Information System (INIS)

Calvo, C.; Saenz de Tejada, L. M.; Diaz Diaz, J.

1969-01-01

A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI 3 and AI 2 O 3 according to the reaction. (Author)

Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

Energy Technology Data Exchange (ETDEWEB)

Mantecón, Javier González; Mattar Neto, Miguel, E-mail: javier.mantecon@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

International Nuclear Information System (INIS)

Mantecón, Javier González; Mattar Neto, Miguel

2017-01-01

The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

Methodological study for management of the generated effluents during MTR-type fuel elements fabrication at IPEN/CNEN-SP plant

International Nuclear Information System (INIS)

Tanzillo Santos, Glaucia Regina

2008-01-01

Full text: The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the main programs of the Institute of Energetic and Nuclear Research of the National Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel -CCN- is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt % 235 U), to supply its IEA-R1 research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the sustainability concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

International Nuclear Information System (INIS)

Ardaneh, Kazem; Zaferanlouei, Salman

2013-01-01

Highlights: ► A model is presented to simulate the reactivity insertion transient in MTR reactors. ► Transient dynamics of IAEA 10 MW MTR type research reactor are evaluated. ► Maximum unprotected reactivity insertion for safe condition is calculated. ► The model predictions are validated with corresponding results in the literature. - Abstract: On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition

Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Matos, J.E.

1996-01-01

At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U 3 Si 2 fuel is about 6.0 g U/cm 3 . The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm 3 and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm 3 . On the other hand, UN is the least reactive fuel because of the relatively large 14 N(n,p) cross section. For a fixed value of k eff , the required 235 U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO 2 dispersions are only useful for uranium densities below 5.0 g/cm 3 . In this density range, however, UO 2 is more reactive than U 3 Si 2

NAC international dry spent fuel transfer technology

International Nuclear Information System (INIS)

Shelton, Thomas A.; Malone, James P.; Patterson, John R.

1996-01-01

Full text: For more than ten years NAC International (NAC) has designed, fabricated, tested and operated a variety of Dry Transfer Systems (DTS's) to transfer spent nuclear fuel from facilities with limited crane capabilities, limited accesses or limiting features to IAEA and USNRC licensed spent fuel transport casks or vice-versa. These DTS's have been operated in diverse environments in the United States and throughout the world and have proven to be a significant enhancement in transferring fuel between spent fuel pools, dry storage and hot cell facilities and spent fuel transport casks. Over the years, NAC has successfully and safely transferred more than two thousand fuel assemblies in DTS's. Our latest generation DTS incorporates years of extensive design and operating experience. It consists of a transfer cask with integrated fuel canister grapple, fuel canisters, and facility and cask adapters as well as a complement of related tools and equipment. The transfer cask is used to move irradiated HEU and LEU MTR fuel onsite in those instances where direct loading or unloading of the shipping cask is not possible due to dimensional, weight or other restrictions. The transfer cask is used to move canisters of fuel from the fuel storage location to the shipping cask. Adapters are employed to ensure proper interfacing of the transfer cask with fuel storage locations and shipping casks (NAC-LWT and NLI-1/2). Our existing fuel storage location adapter is designed for use with a storage pool; however, site or equipment specific adapters can easily be developed to allow interfacing with virtually any storage facility. Prior to movement of the first fuel canister in the transfer cask, the shipping cask is prepared for loading by proper set up of the base plate, shipping cask and shipping cask adapter. The fuel canisters are loaded with fuel and then retracted into the transfer cask via the fuel storage location adapter. The transfer cask is then moved to the shipping

The use of U3Si2 dispersed in aluminum in plate-type fuel elements for research and test reactors

International Nuclear Information System (INIS)

Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

1987-10-01

A high-density fuel based on U 3 Si 2 dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U 3 Si 2 fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U 3 Si 2 particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U 3 Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U 3 Si 2 -aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m 3 is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs

Flow field bipolar plates in a proton exchange membrane fuel cell: Analysis & modeling

International Nuclear Information System (INIS)

Kahraman, Huseyin; Orhan, Mehmet F.

2017-01-01

Highlights: • Covers a comprehensive review of available flow field channel configurations. • Examines the main design considerations and limitations for a flow field network. • Explores the common materials and material properties used for flow field plates. • Presents a case study of step-by-step modeling for an optimum flow field design. - Abstract: This study investigates flow fields and flow field plates (bipolar plates) in proton exchange membrane fuel cells. In this regard, the main design considerations and limitations for a flow field network have been examined, along with a comprehensive review of currently available flow field channel configurations. Also, the common materials and material properties used for flow field plates have been explored. Furthermore, a case study of step-by-step modeling for an optimum flow field design has been presented in-details. Finally, a parametric study has been conducted with respect to many design and performance parameters in a flow field plate.

Fabrication of AA6061-T6 Plate Type Fuel Assembly Using Electron Beam Welding Process

International Nuclear Information System (INIS)

Kim, Soosung; Seo, Kyoungseok; Lee, Donbae; Park, Jongman; Lee, Yoonsang; Lee, Chongtak

2014-01-01

AA6061-T6 aluminum alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW. However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the shrinkage measurement and weld inspection using computed tomography. This study was carried out to determine the suitable welding parameters and to evaluate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory electron beam welding process of the full-sized sample was being developed. Based on this fundamental study, fabrication of the plate-type fuel assembly will be provided for the future Ki-Jang research reactor project

Improvement of visualization efficiency for the nondestructive inspection image of internal defects in plate type nuclear fuel

International Nuclear Information System (INIS)

Park, Seung Kyu; Park, Nak Kyu; Baik, Sung Hoon; Lee, Yoon Sang; Cheong, Yong Moo; Kang, Young June

2012-01-01

Plate type nuclear fuel has been adopted in most research reactors. The production quality of the fuel is a key part for an efficient and stable generation of thermal energy in research reactors. Thus, a nondestructive quality inspection for the internal defects of plate type nuclear fuel is a key process during the production of nuclear fuel for safety insurance. Nondestructive quality inspections based on X rays and ultrasounds have been widely used for the defect detection of plate type nuclear fuel. X ray testing is a simple and fast inspection method, and provides an image in real time as the inspection results. Thus, the testing can be carried out by a non expert field worker. However, it is hard to detect closed type defects that should be detected during the production of plate type nuclear fuel. Ultrasonic testing is a powerful tool to detect internal defects including open type and closed type defects in plate type nuclear fuel. However, the inspection process is complicated because an immersion test should be carried out in a water tank. It is also a time consuming inspection method because area testing to acquire image is based on the scanning of the point by point inspections. Among nondestructive inspection techniques, the techniques based on laser interferometry and infrared thermography have been widely used in the detection of internal defects of plate type composite materials, such as aircraft, automotive etc. While infrared thermography technique (IRT) analyses the thermal behavior of the specimen surface, laser interferometry technique (LIT) analyses the deformation field. Both techniques are useful tools for detection and evaluation of internal defects in composite materials. Especially, the laser interferometry technique can provide the depth information of internal defects. Laser interferometry technique (LIT) is a non contact inspection method faster than thermography. Also, this technique requires less energy than thermography and the

Systems for neutronic, thermohydraulic and shielding calculation in personal computers

International Nuclear Information System (INIS)

Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.

1990-01-01

The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es

Miniaturized polymer electrolyte fuel cell (PEFC) stack using micro structured bipolar plate

Energy Technology Data Exchange (ETDEWEB)

Veziridis, Z; Scherer, G G; Marmy, Ch; Glaus, F [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1999-08-01

In Polymer Electrolyte Fuel Cell (PEFC) technology the reducing of volume and mass of the fuel cell stack and the improvement of catalyst utilization are of great interest. These parameters affect applicability and system cost. In this work we present an alternative way for reducing the stack volume by combining gas distribution and catalytic active area in one plate. Micro machined glassy carbon electrodes serve as support material for the platinum catalyst, as well as gas distributor at the same time. A comparison of these electrodes with conventional platinum-black gas diffusion electrodes under fuel cell conditions shows that the new system is a promising electrode type for enhanced power density and catalyst utilization. (author) 3 figs., 5 refs.

Flow induced deformation and collapse of a thin rectangular plate with application to the Engineering Test Reactor nuclear fuel elements

International Nuclear Information System (INIS)

Davis, C.D.

1981-01-01

This work examines a single flat fuel plate bounded by two channels and determines static plate deflections, resultant forces and bending stresses due to pressure differential and hydrodynamic loadings. The classical then reactangular plate equations are used to model the fuel plate. These equations contain as an input the hydrodynamic loading function for modeling the fluid-structural interaction. Two models of the channel flow are developed. One assumes the accelerating potential core flow is laminar with developing two-dimensional laminar boundary layers being formed on the channel walls. The Schlichting entry length solution for developing laminar flow is found to be valid the entire length of the channel. The second model assumes the core flow is fully-developed turbulent the entire length of the channel. Hydrodynamic loading functions are developed for both flow models containing parameters for fluid density, fluid velocity, Reynolds number and channel and plate dimensions. Hence the effects of each parameter can be examined independently. A criterion is developed for predicting ETR fuel plate collapse at high channel flow velocities, 1067 cm/s (420 in/sec) (R/sub e/ = 60,000). The criterion predicts that in order to prevent ETR plate collapse the inlet velocities between channels must not differ by more than 2%

The determination of uranium distribution homogeneity in the fuel plates with the uranium loading of 4.80 and 5.20 g/cm3 by X-Ray attenuation

International Nuclear Information System (INIS)

Supardjo; Rojak, A.; Boybul; Suyoto; Datam, A. S.

2000-01-01

The calibration of X-Ray intensity of the U 3 Si 2 -AI fuel plates with the uranium loading between 3.60 up to 5.20 g/cm 3 and varied thickness of AIMgSi1 reference block have been performed. The measurement with changing variable slit diameter and energy of X-Ray attenuation, are produced enough representative X-Ray intensity at 18 mm slit diameter and energy of 43 kV. From the correlation of X-ray intensities vs variation of uranium loading in the fuel plates and thickness of the AIMgSi1 materials, the equivalence of thickness of the AIMgSi1 block to the uranium loading of fuel plates are determined. By assuming that the tolerance of the homogeneity measurement is + 20 % from normal thickness staircase of the AIMgSi1 standard could be determined and than together with fuel plate were scanned to determine the uranium homogeneity. The test result on the U 3 Si 2 -AI fuel plates with uranium loading of 4.80 and 5.20 g/cm 3 (each 4 fuel plates) indicated that uranium distribution in the fuel plates is relatively homogeneous, with each maximum deviation being 6.30 % and 6.90%. It is showed that measurement method is relatively good, easy, and fast so that this method is suitable to control the uranium homogeneity in the fuel plate. (author)

Thick nickel plating of spent fuel transport and storage casks CASTOR and POLLUX

International Nuclear Information System (INIS)

Wilbuer, K.

1991-01-01

Spent fuel elements have to be safely handled in containers for transport and storage. These large casks (100-120 t) are made by various firms according to the specifications given by the nuclear plant operator. For shielding and protection of the hazardous material, the casks' inner surface is coated with a nickel plating about 3000 μm thick. The product and the production process are subject to very stringent requirements, due to the hazardous potential of the material to be shipped or stored. Therefore, both the extremely high quality standards to be met by the nickel plating and the dimensions and capability of the plating plant required for the process are problems that cannot be solved by a usual commercial plating plant. The new concept and process that had to be established are explained in the paper. (orig./MM) [de

CERCA's 25 years experience in U3Si2 fuel manufacturing

International Nuclear Information System (INIS)

Durand, JP.; Duban, B.; Lavastre, Y.; Perthuis, S. de

2003-01-01

This paper documents the experience gained at CERCA in manufacturing, testing, and inspecting U 3 Si 2 fuel elements for various Material Test Reactors (MTR) since the beginning of the RERTR Program in 1978, up to now. It emphasises how the company controls the product to insure compliance with the fuel-related safety parameters. Finally, those statements are considered in the UMo fuel production perspective. (author)

Apparatus for transferring nuclear fuel pellets to a plate loader

International Nuclear Information System (INIS)

Huggins, T.B.

1978-01-01

An apparatus is described for transferring nuclear fuel pellets from a grinding machine to a plate loader. It includes a frame, an endless belt fitted to the frame, a control system provided on it for actuating the belt at a preset speed, a V shaped vessel fitted directly above the belt and extending along its length to guide the pellets on the belt and a device to receive the pellets coming from the belt [fr

Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

Energy Technology Data Exchange (ETDEWEB)

Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2015-07-01

In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

International Nuclear Information System (INIS)

Ball, G.; Pond, R.; Hanan, N.; Matos, J.

1995-01-01

This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions

Requirements and testing methods for surfaces of metallic bipolar plates for low-temperature PEM fuel cells

Science.gov (United States)

Jendras, P.; Lötsch, K.; von Unwerth, T.

2017-03-01

To reduce emissions and to substitute combustion engines automotive manufacturers, legislature and first users aspire hydrogen fuel cell vehicles. Up to now the focus of research was set on ensuring functionality and increasing durability of fuel cell components. Therefore, expensive materials were used. Contemporary research and development try to substitute these substances by more cost-effective material combinations. The bipolar plate is a key component with the greatest influence on volume and mass of a fuel cell stack and they have to meet complex requirements. They support bending sensitive components of stack, spread reactants over active cell area and form the electrical contact to another cell. Furthermore, bipolar plates dissipate heat of reaction and separate one cell gastight from the other. Consequently, they need a low interfacial contact resistance (ICR) to the gas diffusion layer, high flexural strength, good thermal conductivity and a high durability. To reduce costs stainless steel is a favoured material for bipolar plates in automotive applications. Steel is characterized by good electrical and thermal conductivity but the acid environment requires a high chemical durability against corrosion as well. On the one hand formation of a passivating oxide layer increasing ICR should be inhibited. On the other hand pitting corrosion leading to increased permeation rate may not occur. Therefore, a suitable substrate lamination combination is wanted. In this study material testing methods for bipolar plates are considered.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

Science.gov (United States)

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Fission rate measurements in fuel plate type assembly reactor cores

International Nuclear Information System (INIS)

Rogers, J.W.

1988-01-01

The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

In-core instrumentation and in-situ measurement in connection with fuel behaviour. Working material

International Nuclear Information System (INIS)

1996-01-01

The subject of this meeting has been touched on briefly in most of the Specialist's and topical meetings related to fuel behaviour. On the basis of the conclusions and recommendations of these meetings the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended the Agency to organize a dedicated Specialist's Meeting on the subject. The twenty one papers covered the instrumentation, sensors, methods and computer codes currently used in Material Test Reactor (MTR) and power reactors as well as improved instrumentation and methods. The meeting acknowledged the fast development of fuel modelling and therefore the growing need of dedicated high burnup fuel experiments carried out in MTR reactors on refabricated rods from power reactors. In order to reduce safety margins in power reactors, thus improving economics, the necessity to develop more sophisticated on-line calculations, based on improved sensors, was recognized, although this development is limited by insufficient knowledge of the mechanisms involved. Refs, figs, tabs

Control of gdhR Expression in Neisseria gonorrhoeae via Autoregulation and a Master Repressor (MtrR of a Drug Efflux Pump Operon

Directory of Open Access Journals (Sweden)

Corinne E. Rouquette-Loughlin

2017-04-01

Full Text Available The MtrCDE efflux pump of Neisseria gonorrhoeae contributes to gonococcal resistance to a number of antibiotics used previously or currently in treatment of gonorrhea, as well as to host-derived antimicrobials that participate in innate defense. Overexpression of the MtrCDE efflux pump increases gonococcal survival and fitness during experimental lower genital tract infection of female mice. Transcription of mtrCDE can be repressed by the DNA-binding protein MtrR, which also acts as a global regulator of genes involved in important metabolic, physiologic, or regulatory processes. Here, we investigated whether a gene downstream of mtrCDE, previously annotated gdhR in Neisseria meningitidis, is a target for regulation by MtrR. In meningococci, GdhR serves as a regulator of genes involved in glucose catabolism, amino acid transport, and biosynthesis, including gdhA, which encodes an l-glutamate dehydrogenase and is located next to gdhR but is transcriptionally divergent. We report here that in N. gonorrhoeae, expression of gdhR is subject to autoregulation by GdhR and direct repression by MtrR. Importantly, loss of GdhR significantly increased gonococcal fitness compared to a complemented mutant strain during experimental murine infection. Interestingly, loss of GdhR did not influence expression of gdhA, as reported for meningococci. This variance is most likely due to differences in promoter localization and utilization between gonococci and meningococci. We propose that transcriptional control of gonococcal genes through the action of MtrR and GdhR contributes to fitness of N. gonorrhoeae during infection.

Surface roughness effect on the metallic bipolar plates of a proton exchange membrane fuel cell

International Nuclear Information System (INIS)

Lin, Chien-Hung

2013-01-01

Highlights: ► Various degrees of roughness are caused by the sandblasting method. ► An improper surface modification depletes the PEMFC performance severely. ► The AC impedance are used to assess the fuel gas transfer effect. ► The Warburg resistance form in the coarse flow channel surface. - Abstract: Proton exchange membrane fuel cells (PEMFCs) is a promising candidate as energy systems. However, the stability and lifetime of cells are still important issues. The effect of surface roughness on metallic bipolar plate is discussed in this paper. Various roughness on the bulk surface are obtained by the sandblasting method. The grain sizes of sand are selected as 50, 100 and 200 μm. The Ac impedance experiment results show that the bipolar plate roughness and carbon paper porosity are well matched when the surface roughness is within 1–2 μm. Superior condition decreases the contact resistance loss in the fuel cell. The high frequency resistance of the coarse surface was larger than that of the substrate by around 5 mΩ. Furthermore, a new arc was formed at the low frequency region. Hence, the unmatch roughness condition of the bipolar plate significantly increases the contact resistance and mass transfer resistance. This paper develops a sequential approach to study an optimum surface roughness by combining the whole performance (I–V) curve and AC impedance result. It benefits us to quantify the contact and mass transfer resistance exists in the PEMFC. The proposed surface treatment improves the surface effect and promotes the implement of potential metallic bipolar plate in near future

Validation concerns for dry storage of foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Trumble, E.F.

1994-01-01

Recent decisions by the Department of Energy have accelerated the need for storage options to support the return of foreign research reactor (FRR) fuel to the United States. Many of these returns consist of fuel types which contain highly enriched uranium and are aluminum clad. These attributes present many challenges not experienced in the fuel storage designs for commercial nuclear fuels where the fuels have lower enrichment and the cladding is more robust. Historically, returned FRR fuel has been stored for short periods in basins where it is cooled and then sent to be reprocessed. However, a severe lack of basin space and questionable availability of reprocessing facilities necessitates the development of other proposals. One proposed option is to store the FRR fuel in a dry state, thus reducing the corrosion problems associated with aluminum cladding. A drawback to this type of storage, however, is the lack of experimental data for this type of fuel under dry storage conditions. This lack of data has led to recent discussions over the accuracy of some of the current multigroup cross section libraries when applied to dry, fast systems of uranium and aluminum. This concern is evaluated for the specific case of Material Test Reactor (MTR) fuel (MTR is >60% of FRR fuel), a review of applicable experiments is presented and a new experiment is proposed

A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

International Nuclear Information System (INIS)

Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

2016-01-01

The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature

Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

Directory of Open Access Journals (Sweden)

Sidi Ali Kamel

2012-01-01

Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses

International Nuclear Information System (INIS)

Rios, Ilka Antonia

2013-01-01

This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

Structure and reconstitution of yeast Mpp6-nuclear exosome complexes reveals that Mpp6 stimulates RNA decay and recruits the Mtr4 helicase

Energy Technology Data Exchange (ETDEWEB)

Wasmuth, Elizabeth V. [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Zinder, John C. [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Tri-Institutional Training Program in Chemical Biology, Memorial Sloan Kettering Cancer Center, New York, United States; Zattas, Dimitrios [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Das, Mom [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Lima, Christopher D. [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Howard Hughes Medical Institute, Memorial Sloan Kettering Cancer Center, New York, United States

2017-07-25

Nuclear RNA exosomes catalyze a range of RNA processing and decay activities that are coordinated in part by cofactors, including Mpp6, Rrp47, and the Mtr4 RNA helicase. Mpp6 interacts with the nine-subunit exosome core, while Rrp47 stabilizes the exoribonuclease Rrp6 and recruits Mtr4, but it is less clear if these cofactors work together. Using biochemistry with Saccharomyces cerevisiae proteins, we show that Rrp47 and Mpp6 stimulate exosome-mediated RNA decay, albeit with unique dependencies on elements within the nuclear exosome. Mpp6-exosomes can recruit Mtr4, while Mpp6 and Rrp47 each contribute to Mtr4-dependent RNA decay, with maximal Mtr4-dependent decay observed with both cofactors. The 3.3 Å structure of a twelve-subunit nuclear Mpp6 exosome bound to RNA shows the central region of Mpp6 bound to the exosome core, positioning its Mtr4 recruitment domain next to Rrp6 and the exosome central channel. Genetic analysis reveals interactions that are largely consistent with our model.

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

Directory of Open Access Journals (Sweden)

Itamar Iliuk

2016-01-01

Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

Structural analysis of hatch cover plates on Fuels and Materials Examination Facility high bay mezzanine

International Nuclear Information System (INIS)

Dixson, G.E.

1997-01-01

In order to move the Idaho National Engineering Laboratory (INEL) Light Duty Utility Arm (LDUA) trailer into position for testing on the Fuels and Materials Examination Facility (FMEF) 42 ft level mezzanine one of the trailer's wheels will have to sit on a circular hatch cover fabricated from one-inch thick steel plate. The attached calculations verify that the hatch cover plate is strong enough to support the weight of the INEL LDUA trailer's wheel

Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

Energy Technology Data Exchange (ETDEWEB)

Borring, J.; Gundtoft, H.E.; Borum, K.K.; Toft, P. [Riso National Lab. (Denmark)

1997-08-01

In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptance of a thinner nominal cladding than normally used today.

Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

International Nuclear Information System (INIS)

Borring, J.; Gundtoft, H.E.; Borum, K.K.; Toft, P.

1997-01-01

In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptance of a thinner nominal cladding than normally used today

Economical analysis to utilize MTR fuel elements using silicides in research reactors

International Nuclear Information System (INIS)

Bergallo, Juan E.; Novara, Oscar E.; Adelfang, Pablo

2000-01-01

According to international programs on reducing enrichment in research reactors and the necessity to maintain their operation, new fuel elements have been developed in order to meet both objectives. Thus, U-Si alloy fuel elements for research reactors are becoming of greater interest for the international markets. It became necessary to make an economic study about the convenience of introducing this type of fuel elements in the RA-3 reactor and to know the potentiality of this fuel. The economical behavior of the reactor operation has been evaluated comparing the actual U 3 O 8 nuclear fuel cycle with U 3 Si 2 nuclear fuels. Results obtained show that the main economical factor to determine the change of fuels is the cost of fabrication, and the change is advisable up to an 80% difference. The other factors related to the cost of nuclear fuel cycle are not relevant or have real minor impacts. (author)

Heat conduction in a plate-type fuel element with time-dependent boundary conditions

International Nuclear Information System (INIS)

Faya, A.J.G.; Maiorino, J.R.

1981-01-01

A method for the solution of boundary-value problems with variable boundary conditions is applied to solve a heat conduction problem in a plate-type fuel element with time dependent film coefficient. The numerical results show the feasibility of the method in the solution of this class of problems. (Author) [pt

Multi-target retrieval (MTR): the simultaneous retrieval of pressure, temperature and volume mixing ratio profiles from limb-scanning atmospheric measurements

International Nuclear Information System (INIS)

Dinelli, B.M.; Alpaslan, D.; Carlotti, M.; Magnani, L.; Ridolfi, M.

2004-01-01

In this paper we describe a retrieval approach for the simultaneous determination of the altitude distributions of p, T and VMR of atmospheric constituents from limb-scanning measurements of the atmosphere. This analysis method, named multi-target retrieval (MTR), has been designed and implemented in a computer code aimed at the analysis of MIPAS-ENVISAT observations; however, the concepts implemented in MTR have a general validity and can be extended to the analysis of all type of limb-scanning observations. In order to assess performance and advantages of the proposed approach, MTR has been compared with the sequential analysis system implemented by ESA as the level-2 processor for MIPAS measurements. The comparison has been performed on a common set of target species and spectral intervals. The performed tests have shown that MTR produces results of better quality than a sequential retrieval. However, the simultaneous retrieval of p, T and water VMR has not lead to satisfactory results below the tropopause, because of the high correlation occurring between p and water VMR in the troposphere. We have shown that this problem can be fixed extending the MTR analysis to at least one further target whose spectral features decouple the retrieval of pressure and water VMR. Ozone was found to be a suitable target for this purpose. The advantages of the MTR analysis system in terms of systematic errors have also been discussed

Status of fuel element technology for plate type dispersion fuels with high uranium density

International Nuclear Information System (INIS)

Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

1983-01-01

A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

High loading uranium plate

International Nuclear Information System (INIS)

Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

1990-01-01

Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process

Suitability of x-ray paper as an inspection tool for flat plate nuclear fuel

International Nuclear Information System (INIS)

Barna, B.A.

1979-01-01

The flat plate nuclear fuel used in the Advanced Test Reactor (ATR) has several attributes which are best examined by radiography. These are fuel core dimensions and location, homogeneity of the uranium aluminide alloy that composes the core, and the location and sizing of fuel particles in the fuel free edge borders of the plates. The most economiccal approach is to inspect for all three attributes from a single radiograph which requires accommodation of a large contrast range. Currently radiography is conducted using Kodak type M double emulsion film which provides a high quality image for evaluation. A promising alternative to film exists however in paper radiography. The two media are very similar except that paper uses a single emulsion which is deposited on an opaque diffuse reflecting surface. This requires that the image be viewed with reflected rather than transmitted light. This type of physical structure results in lower materials and processing costs. For example, Kodak Industrex 600 paper is approximately 50% the cost of type M film. In addition the image can be developed and viewed (although not fixed) in as little as 10 seconds. The results of test to ascertain the suitability of paper radiography for these purposes are described. Whole there was some degradation of the image with the use of paper, the paper was judged suitable for identification of edge border location, homogeneity, and floking

The technique for determination of surface contamination by uranium on U3Si2-Al plate-type fuel elements

International Nuclear Information System (INIS)

Li Shulan; He Fengqi; Wang Qingheng; Han Jingquan

1993-04-01

The NDT method for determining the surface contamination by uranium on U 3 Si 2 -Al plate-type fuel elements, the process of standard specimen preparation and the graduation curve are described. The measurement results of U 3 Si 2 -Al plate-type fuel elements show that the alpha counting method to measure the surface contamination by uranium on fuel plate is more reliable. The UB-1 type surface contamination meter, which was recently developed, has many advantages such as high sensitivity to determine the uranium pollution, short time in measuring, convenience for operation, and the minimum detectable amount of uranium is 5 x 10 -10 g/cm 2 . The measuring device is controlled by a microcomputer. Besides data acquisition and processing, it has functions of statistics, output data on terminal or to printer and alarm. The procedures of measurement are fully automatic. All of these will meet the measuring needs in batch process

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

Energy Technology Data Exchange (ETDEWEB)

Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

2018-02-01

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Sicoated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, is temperature and fission-rate dependent. In order to simulate the U-Mo/Al inter-diffusion layer (IL) growth behavior in full-size dispersion fuel plates, the existing IL growth correlation was modified with a temperaturedependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the intermixing rate in ion-irradiated bi-layer systems.

Effect of temperature on the expansion and microstructure Of U3 Si2-AI mini plate fuel of 3.6 g/cm3 uranium loading

International Nuclear Information System (INIS)

Ginting, A. Br.; Samosir, N.; Suparjo; Nasution, H.

2000-01-01

Expansion analysis has been conducted to 50 x 20-mm U 3 Si 2 -AI mini plate of 3.6 g/cm 3 uranium loading using dilatometer. The analysis was carried out at various temperatures of 170 o C, 350 o C and 550 o C in Argon medium with delay time 4 days. The result showed that the fuel plate was relatively stable with increasing of heating time but underwent significant expansion. Heating at 170 o C, 350 o C and 550 o C resulted in the expansion of the U 3 Si 2 -AI fuel plate of to 83-212 mum, 333-475 mum, and 433-724 mum with coefficient expansion of 24.2x10 -6 / o C - 24.3x10 -6 / o C, 25.5x10 -6 / o C - 26.2x10 -6 /'oC and 26.6 x 10 -6 / o C - 28.2 x 10 -6 / o C respectively. Microanalysis of the U 3 Si 2 -AI mini plate fuel with SEM-EDS upon heating at those temperature variation showed that microstructure change didn't occur at 170 o C, mean while interaction between AIMg2 cladding and the fuel meat appeared to take place at 350 o C and 550 o C. Data on the expansion and microstructure change of U 3 Si 2 -AI fuel plate upon heating are of great important for the manufacture/fabrication of research fuel plate to produce silicide fuel element for higher uranium loading. (author)

Evaluation of Electron Beam Welding Performance of AA6061-T6 Plate-type Fuel Assembly

International Nuclear Information System (INIS)

Kim, Soo-Sung; Seo, Kyoung-Seok; Lee, Don-Bae; Park, Jong-Man; Lee, Yoon-Sang; Lee, Chong-Tak

2014-01-01

As one of the most commonly used heat-treatable aluminum alloys, AA6061-T6 aluminum alloy is available in a wide range of structural materials. Typically, it is used in structural members, auto-body sheet and many other applications. Generally, this alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW(Electron Beam Welding). However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the plate-type nuclear fuel fabrication and assembly, a fundamental electron beam welding experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the suitable welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the plate-type fuel assembly has been also studied by the weld penetrations of side plate to end fitting and fixing bar and weld inspections using computed tomography

Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

International Nuclear Information System (INIS)

Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

2013-01-01

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

Crystal structure of the Neisseria gonorrhoeae MtrD inner membrane multidrug efflux pump.

Directory of Open Access Journals (Sweden)

Jani Reddy Bolla

Full Text Available Neisseria gonorrhoeae is an obligate human pathogen and the causative agent of the sexually-transmitted disease gonorrhea. The control of this disease has been compromised by the increasing proportion of infections due to antibiotic-resistant strains, which are growing at an alarming rate. The MtrCDE tripartite multidrug efflux pump, belonging to the hydrophobic and amphiphilic efflux resistance-nodulation-cell division (HAE-RND family, spans both the inner and outer membranes of N. gonorrhoeae and confers resistance to a variety of antibiotics and toxic compounds. We here report the crystal structure of the inner membrane MtrD multidrug efflux pump, which reveals a novel structural feature that is not found in other RND efflux pumps.

Scanning electron microscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices

Energy Technology Data Exchange (ETDEWEB)

Keiser, Dennis D. Jr; Jue, Jan Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adom B.; Medvedev, Pavel; Madden, James; Wachs, Dan; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory (United States)

2014-04-15

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifically, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (-4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

Design and performance of tubular flat-plate solid oxide fuel cell

Energy Technology Data Exchange (ETDEWEB)

Matsushima, T.; Ikeda, D.; Kanagawa, H. [NTT Integrated Information & Energy Systems Labs., Tokyo (Japan)] [and others

1996-12-31

With the growing interest in conserving the environmental conditions, much attention is being paid to Solid Oxide Fuel Cell (SOFC), which has high energy-conversion efficiency. Many organizations have conducted studies on tubular and flat type SOFCs. Nippon Telegraph and Telephone Corporation (NTT) has studied a combined tubular flat-plate SOFC, and already presented the I-V characteristics of a single cell. Here, we report the construction of a stack of this SOFC cell and successful generation tests results.

An experimental investigation of the interaction of primary and secondary stresses in fuel plates

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1996-01-01

If the load is not relieved as a structure starts to yield, the induced stress is defined as primary stress. If the load relaxes, as a structure begins yield the induced stress is defined as secondary stress. In design it is not uncommon to give more weight to primary stresses than to secondary stresses. However, knowing when this is good design practice and when it is not good design practice represents a problem. In particular, the fuel plates in operating reactors contain both primary stresses and secondary stresses and to properly assess a design there is a need to assign design weights to the stresses. Tests were conducted on reactor fuel plates intended for the Advanced Neutron Source (ANS) to determine the potential of giving different design weights to the primary and secondary stresses. The results of these tests and the conclusion that the stresses should be weighted the same are given in this paper

An experimental investigation on the interaction of primary and secondary stresses in fuel plates

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1997-01-01

If the load is not relieved as a structure starts to yield, the induced stress is defined as primary stress. If the load relaxes, as a structure begins to yield the induced stress is defined as secondary stress. In design, it is not uncommon to give more weight to primary stresses than to secondary stresses. However, knowing when this is good design practice and when it is not good design practice represent a problem. In particular, the fuel plates in operating reactors contain both primary stresses and secondary stresses, and to properly assess a design there is a need to assign design weights to the stresses. Tests were conducted on reactor fuel plates intended for the advanced neutron source (ANS) to determine the potential of giving different design weights to the primary and secondary stresses. The results of these tests and the conclusion that the stresses should be weighted the same are given in this paper

Static analysis of material testing reactor cores:critical core calculations

International Nuclear Information System (INIS)

Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

1999-01-01

A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

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

Directory of Open Access Journals (Sweden)

H. Rashtchi

2018-03-01

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

C.E.R.C.A. contribution to the RERTR program. Status of development, November 1982

International Nuclear Information System (INIS)

Savornin, B.; Fanjas, Y.R.

1983-01-01

For more than 20 years, CERCA has been manufacturing MTR fuel plates and fuel elements. 200.000 fuel plates have thus been produced and irradiated in reactors of 15 countries through the world, including the U.S.A. and Japan. On the way to the use of LEU in research reactors, we have adapted our fuel plate fabrication technology and developed new fuels in order to achieve high uranium densities in the meat. These technological developments have been performed on our own funds. Aluminide (U Al x ), Oxide (U 3 O 8 ) and Silicide (U 3 Si, U 3 Si 2 , U 3 Si Al) fuels have been successfully developed. Upon completion of the aluminide and oxide developments, we have intensified our effort on silicide fuels. After a brief review of the technological results obtained with U Al x and U 3 O 8 materials, silicide fuels results are examined in more detail. The irradiation status of these various fuels is discussed. It has been shown that CERCA is in position to provide on industrial scale UAl x , U 3 O 8 and U 3 Si 2 fuels with uranium densities up to 5 g/cm 3 . New equipment has been installed in MTR workshop in ROMANS for this purpose. For densities between and 74 g U/cm 3 , pilot technology is available. Development is carried out towards industrialization of U 3 Si fuel production. The commercial availability of these fuels depends on their qualification by post irradiation examination. It can be seen that aluminide fuel will be commercially available in 1983, oxide fuel in 1984. As far as silicide fuels are concerned, their availability will depend on the number of tests to be performed and the burn-up levels required by the Safety Commissions to consider the fuel qualified. These requirements may be different from one reactor to the other. This is why the corresponding dates stretch from 1985 to 1988. The goal of technological development has been reached. Densities above 7 g U/cm 3 have been achieved. Assuming silicide fuels reprocessing does not present major

Analysis of the effect of transverse power distribution in an involute fuel plate with and without oxide film formation

International Nuclear Information System (INIS)

Smith, R. S.

1998-01-01

Existing thermal hydraulics computer codes can account for variations in power and temperature in the axial and thickness directions but variations across the width of the plate cannot be accounted for. In the case of fuel plates in an annular core this can lead to significant errors which are accentuated by the presence of an oxide layer that builds up on the aluminum cladding with burnup. This paper uses a three dimensional SINDA model to account for the transverse variations in power. The effect of oxide thickness on these differences is studied in detail. Power distribution and fuel conductivity are also considered. The lower temperatures predicted with the SINDA model result in a greater margin to clad and fuel damage

Characterization of Thermal and Mechanical Properties of Polypropylene-Based Composites for Fuel Cell Bipolar Plates and Development of Educational Tools in Hydrogen and Fuel Cell Technologies

Science.gov (United States)

Lopez Gaxiola, Daniel

2011-01-01

In this project we developed conductive thermoplastic resins by adding varying amounts of three different carbon fillers: carbon black (CB), synthetic graphite (SG) and multi-walled carbon nanotubes (CNT) to a polypropylene matrix for application as fuel cell bipolar plates. This component of fuel cells provides mechanical support to the stack,…

Nuclear fuel management in JMTR

International Nuclear Information System (INIS)

Naka, Michihiro; Miyazawa, Masataka; Sato, Hiroshi; Nakayama, Fusao; Ito, Haruhiko

1999-01-01

The Japan Materials Testing Reactor (JMTR) is the largest scale materials (author)ted the fission gas release compared with the steady state opkW/l in Japan. JMTR as a multi-purpose reactor has been contributing to research and development on nuclear field with a wide variety of irradiation for performing engineering tests and safety research on fuel and component for light water reactor as well as fast breeder reactor, high temperature gas-cooled reactor etc., for research and development on blanket material for fusion reactor, for fundamental research, and for radio-isotope (RI) production. The driver nuclear fuel used in JMTR is aluminum based MTR type fuel. According to the Reduced Enrichment for Research and Test Reactors (RERTR) Program, the JMTR fuel elements had been converted from 93% high enriched uranium (HEU) fuel to 45% medium enriched uranium (MEU) fuel in 1986, and then to 20% low enriched uranium (LEU) fuel in 1994. The cumulative operation cycles until March 1999 reached to 127 cycles since the first criticality in 1968. JMTR has used 1,628 HEU, 688 MEU and 308 LEU fuel elements for these operation cycles. After these spent fuel elements were cooled in the JMTR water canal more than one year after discharged from the JMTR core, they had been transported to reprocessing plants in Europe, and then to plants in USA in order to extract the uranium remaining in the spent fuel. The JMTR spent fuel transportation for reprocessing had been continued until the end of 1988. However, USA had ceased spent fuel reprocessing in 1989, while USDOE committed to prepare an environmental review of the impacts of accepting spent fuels from foreign research reactors. After that, USDOE decided to implement a new acceptance policy in 1996, the spent fuel transportation from JMTR to Savannah River Site was commenced in 1997. It was the first transportation not only in Japan but in Asia also. Until resuming the transportation, the spent fuel elements stored in JMTR

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

International Nuclear Information System (INIS)

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

2013-01-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

Fuel assembly

International Nuclear Information System (INIS)

Fujibayashi, Toru.

1970-01-01

Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

Fuel assembly

International Nuclear Information System (INIS)

Sakuyama, Tadashi; Mukai, Hideyuki.

1988-01-01

Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

Spacer grid for fuel assembly of nuclear reactor comprising opposite support points made with elastic thin plates

International Nuclear Information System (INIS)

Feutrel, C.

1983-01-01

Two series of thin walls form square cells, each containing a fuel pencil. Support points are made in the cells walls. Splines obtained by two parallel slots in the length of the cells. The reaction of fuel pencil produce a deformation of the elastic splines made in the plate, for compensation of the tolerance allowed on the diameter of the pencils [fr

PWR integral tie plate and locking mechanism

International Nuclear Information System (INIS)

Flora, B.S.; Osborne, J.L.

1980-01-01

A locking mechanism for securing an upper tie plate to the tie rods of a nuclear fuel bundle is described. The mechanism includes an upper tie plate assembly and locking sleeves fixed to the ends of the tie rods. The tie plate is part of the upper tie plate assembly and is secured to the fuel bundle by securing the entire upper tie plate assembly to the locking sleeves fixed to the tie rods. The assembly includes, in addition to the tie plate, locking nuts for engaging the locking sleeves, retaining sleeves to operably connect the locking nuts to the assembly, a spring biased reaction plate to restrain the locking nuts in the locked position and a means to facilitate the removal of the entire assembly as a unit from the fuel bundle

Fundamental Study of Electron Beam Welding of AA6061-T6 Aluminum Alloy for Nuclear Fuel Plate Assembly (II)

International Nuclear Information System (INIS)

Kim, Soosung; Lee, Haein; Lee, Donbae; Park, Jongman; Lee, Yoonsang

2013-01-01

Certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes posses the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using a electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. In this experiment, a feasibility test was carried out by tensile tester, bead-on-plate welding and metallographic examination to comply with the aluminum welding procedure. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the mechanical testing and microstructure examinations. This study was carried out to determine the suitable welding process and to investigate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory EBW of the square butt weld specimens was developed. In comparison with the rolling directions of test specimens, the tensile strengths were no difference between the longitudinal and transverse welds. Based on this fundamental study, fabrication and assembly of the nuclear fuel plates will be provided for the future Kijang research reactor project

Post-irradiation examination of U3SIX-AL fuel element manufactured and irradiated in Argentina

International Nuclear Information System (INIS)

Ruggirello, Gabriel; Calabroni, Hector; Sanchez, Miguel; Hofman, Gerard

2002-01-01

As a part of CNEA's qualification program as a supplier of low enriched Al-U 3 Si 2 dispersion fuel elements for research reactors, a post irradiation examination (PIE) of the first prototype of this kind, called P-04, manufactured and irradiated in Argentina, was carried out. The main purpose of this work was to set up various standard PIE techniques in the hot cell, looking forward to the next steps of the qualification program, as well as to acquire experience on the behaviour of this nuclear material and on the control of the manufacturing process. After an appropriate cooling period, on May 2000 the P-04 was transported to the hot cell in Ezeiza Atomic Centre. Non destructive and destructive tests were performed following the PIE procedures developed in Argonne National Laboratory (ANL), this mainly included dimensional measurement, microstructural observations and chemical burn-up analyses. The methodology and results of which are outlined in this report. The results obtained show a behaviour consistent with that of other fuel elements of the same kind, tested previously. On the other hand the results of this PIE, specially those concerning burn-up analysis and stability and corrosion behaviour of the fuel plates, will be of use for the IAEA Regional Program on the characterization of MTR spent fuel. (author)

Flat plate bonded fuel elements. Quarterly report No. 3, October 11, 1953--December 10, 1953

Energy Technology Data Exchange (ETDEWEB)

NONE

1953-12-31

This document is Report No. 3 (covering the period 10/11/53 to 12/10/53) on Flat Plate Bonded Fuel Elements at the Savannah River Plant. It contains information on the fabrication and testing of the uranium components as well as the structural components (aluminium).

Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

International Nuclear Information System (INIS)

Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

2006-01-01

Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

Evaluation of Erosion of the Dummy ''EE'' Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

International Nuclear Information System (INIS)

Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.; Rezvoi, Aleksey V.

2016-01-01

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and ''horseshoeing'' defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum ''dummy'' plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth ''S'' curve, was represented by a series temperature rise ''humps,'' which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping and

Preliminary experience and near future utilization programmes of the MPR-30 fueled by LEU [low enriched uranium

International Nuclear Information System (INIS)

Arbie, B.; Soentono, S.

1987-01-01

The MTR type reactor MPR-30 G.A. Siwabessy, located at PUSPIPTEK Serpong has recently reached its first criticality. This multipurpose reactor is supposed to be the first MTR type reactor in the world that is designed and constructed to be fueled by low enriched uranium. Preliminary experience covering the approach to the first criticality and the excess reactivity loading as well as some thermal hydraulics and power ascension tests are briefly presented and discussed. The near future utilization programmes during and after commissioning are also presented. (Author)

Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

International Nuclear Information System (INIS)

Ding Shurong; Wang Qiming; Huo Yongzhong

2010-01-01

In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo-elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

Anticorrosion Coating of Carbon Nanotube/Polytetrafluoroethylene Composite Film on the Stainless Steel Bipolar Plate for Proton Exchange Membrane Fuel Cells

Directory of Open Access Journals (Sweden)

Yoshiyuki Show

2013-01-01

Full Text Available Composite film of carbon nanotube (CNT and polytetrafluoroethylene (PTFE was formed from dispersion fluids of CNT and PTFE. The composite film showed high electrical conductivity in the range of 0.1–13 S/cm and hydrophobic nature. This composite film was applied to stainless steel (SS bipolar plates of the proton exchange membrane fuel cell (PEMFC as anticorrosion film. This coating decreased the contact resistance between the surface of the bipolar plate and the membrane electrode assembly (MEA of the PEMFC. The output power of the fuel cell is increased by 1.6 times because the decrease in the contact resistance decreases the series resistance of the PEMFC. Moreover, the coating of this composite film protects the bipolar plate from the surface corrosion.

Development and implementation of computational geometric model for simulation of plate type fuel fabrication process with microspheres dispersed in metallic matrix

International Nuclear Information System (INIS)

Lage, Aldo M.F.; Reis, Sergio C.; Braga, Daniel M.; Santos, Armindo; Ferraz, Wilmar B.

2005-01-01

In this report it is presented the development of a geometric model to simulate the plate type fuel fabrication process with fuels microspheres dispersed in metallic matrix, as well as its software implementation. The developed geometric model encloses the steps of pellets pressing and sintering, as well as the plate rolling passes. The model permits the simulation of structures, where the values of the various variables of the fabrication processes can be studied and modified. The following variables were analyzed: microspheres diameters, density of the powder/microspheres mixing, microspheres density, fuel volume fraction, sintering densification, and rolling passes number. In the model implementation, which was codified in DELPHI programming language, systems of structured analysis techniques were utilized. The structures simulated were visualized utilizing the AutoCAD applicative, what permitted to obtain planes sections in diverse directions. The objective of this model is to enable the analysis of the simulated structures and supply information that can help in the improvement of the dispersion microspheres fuel plates fabrication process, now in development at CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) in cooperation with the CTMSP (Centro Tecnologico da Marinha em Sao Paulo). (author)

Fission product phases in irradiated carbide fuels

International Nuclear Information System (INIS)

Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

1975-09-01

Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

International Nuclear Information System (INIS)

Medvedev, Pavel G.

2009-01-01

Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

Concepts for the interim storage of spent fuel elements from research reactors in the Federal Republic of Germany

International Nuclear Information System (INIS)

Niephaus, D.; Bensch, D.; Quaassdorff, P.; Plaetzer, S.

1997-01-01

Research reactors have been operated in the Federal Republic of Germany since the late fifties. These are Material Test Reactors (MTR) and training, Research and Isotope Facilities of General Atomic (TRIGA). A total of seven research reactors, i.e. three TRIGA and four MTR facilities were still in operation at the beginning of 1996. Provisions to apply to the back-end of the fuel cycle are required for their continued operation and for already decommissioned plants. This was ensured until the end of the eighties by the reprocessing of spent fuel elements abroad. In view of impeding uncertainties in connection with waste management through reprocessing abroad, the development of a national back-end fuel cycle concept was commissioned by the Federal Minister of Education, Science, Research and Technology in early 1990. Development work was oriented along the lines of the disposal concept for irradiated light-water reactor fuel elements from nuclear power plants. Analogously, the fuel elements from research reactors are to be interim-stored on a long-term basis in adequately designed transport and storage casks and then be directly finally disposed without reprocessing after up to forty years of interim storage. As a first step in the development of a concept for interim storage, several sites with nuclear infrastructure were examined and assessed with respect to their suitability for interim storage. A reasonably feasible reference concept for storing the research reactor fuel elements in CASTOR MTR 2 transport and storage casks at the Ahaus interim storage facility (BZA) was evaluated and the hot cell facility and AVR store of Forschungszentrum Juelich (KFA) were proposed as an optional contingency concept for casks that cannot be repaired at Ahaus. Development work was continued with detailed studies on these two conceptual variants and the results are presented in this paper. (author)

Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic U–Mo/Zr fuel plates under a heterogeneous irradiation condition

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yunmei; Gong, Xin; Ding, Shurong, E-mail: dsr1971@163.com

2015-04-15

Highlights: • The three-dimensional stress update algorithms in a co-rotational framework are developed for U–Mo and Zircalloy with the irradiation effects. • An effective method for three-dimensional modeling of the in-pile behaviors in heterogeneously irradiated monolithic fuel plates is established and validated. • The effects of the fission-induced creep effects in the U–Mo foil are investigated in detail. • A deformation phenomenon similar to the irradiation experimental results is obtained. - Abstract: For monolithic fuel plates with U–Mo foil and Zircalloy cladding, the three-dimensional large deformation incremental constitutive relations and stress update algorithms in the co-rotational coordinate framework are developed for the fuel and cladding with their respective irradiation effects involved. Three-dimensional finite element simulation of their in-pile thermo-mechanical coupling behaviors under a location-dependent irradiation condition is implemented via the validated user-defined subroutines UMATHT and UMAT in ABAQUS. Comparison of the simulation results for two cases with or without creep considered in the U–Mo foil indicates that with the irradiation creep included (1) considerable stress-relaxation appears in the U–Mo foil, and the mechanical interaction between fuel and cladding is weakened; (2) approximately identical thickness increments in the plate and fuel foil exist and become comparably larger; (3) plastic deformation in the cladding is significantly diminished.

Experience in producing LEU fuel elements for the RSG-GAS

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1991-01-01

To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

Coating Thickness Measurement of the Simulated TRISO-Coated Fuel Particles using an Image Plate and a High Resolution Scanner

International Nuclear Information System (INIS)

Kim, Woong Ki; Kim, Yeon Ku; Jeong, Kyung Chai; Lee, Young Woo; Kim, Bong Goo; Eom, Sung Ho; Kim, Young Min; Yeo, Sung Hwan; Cho, Moon Sung

2014-01-01

In this study, the thickness of the coating layers of 196 coated particles was measured using an Image Plate detector, high resolution scanner and digital image processing techniques. The experimental results are as follows. - An X-ray image was acquired for 196 simulated TRISO-coated fuel particles with ZrO 2 kernel using an Image Plate with high resolution in a reduced amount of time. - We could observe clear boundaries between coating layers for 196 particles. - The geometric distortion error was compensated for the calculation. - The coating thickness of the TRISO-coated fuel particles can be nondestructively measured using X-ray radiography and digital image processing technology. - We can increase the number of TRISO-coated particles to be inspected by increasing the number of Image Plate detectors. A TRISO-coated fuel particle for an HTGR (high temperature gas-cooled reactor) is composed of a nuclear fuel kernel and outer coating layers. The coating layers consist of buffer PyC (pyrolytic carbon), inner PyC (I-PyC), SiC, and outer PyC (O-PyC) layer. The coating thickness is measured to evaluate the soundness of the coating layers. X-ray radiography is one of the nondestructive alternatives for measuring the coating thickness without generating a radioactive waste. Several billion particles are subject to be loaded in a reactor. A lot of sample particles should be tested as much as possible. The acquired X-ray images for the measurement of coating thickness have included a small number of particles because of the restricted resolution and size of the X-ray detector. We tried to test many particles for an X-ray exposure to reduce the measurement time. In this experiment, an X-ray image was acquired for 196 simulated TRISO-coated fuel particles using an image plate and high resolution scanner with a pixel size of 25Χ25 μm 2 . The coating thickness for the particles could be measured on the image

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.)

Application of nonlinear nodal diffusion method for a small research reactor

International Nuclear Information System (INIS)

Jaradat, Mustafa K.; Alawneh, Luay M.; Park, Chang Je; Lee, Byungchul

2014-01-01

Highlights: • We applied nonlinear unified nodal method for 10 MW IAEA MTR benchmark problem. • TRITION–NEWT system was used to obtain two-group burnup dependent cross sections. • The criticality and power distribution compared with reference (IAEA-TECDOC-233). • Comparison between different fuel materials was conducted. • Satisfactory results were provided using UNM for MTR core calculations. - Abstract: Nodal diffusion methods are usually used for LWR calculations and rarely used for research reactor calculations. A unified nodal method with an implementation of the coarse mesh finite difference acceleration was developed for use in plate type research reactor calculations. It was validated for two PWR benchmark problems and then applied for IAEA MTR benchmark problem for static calculations to check the validity and accuracy of the method. This work was conducted to investigate the unified nodal method capability to treat material testing reactor cores. A 10 MW research reactor core is considered with three calculation cases for low enriched uranium fuel depending on the core burnup status of fresh, beginning-of-life, and end-of-life cores. The validation work included criticality calculations, flux distribution, and power distribution; in addition, a comparison between different fuel materials with the same uranium content was conducted. The homogenized two-group cross sections were generated using the TRITON–NEWT system. The results were compared with a reference, which was taken from IAEA-TECDOC-233. The unified nodal method provides satisfactory results for an all-rod out case, and the three-dimensional, two-group diffusion model can be considered accurate enough for MTR core calculations

Preparation of High-Density Uranium-Silicide U3Sl2-Uss: Effects of Preirradiation Heat Treatment on As-Cast Ingot Fuel Plates

International Nuclear Information System (INIS)

Suripto, A; Yuwono

1998-01-01

Heat treatment experiments upon U 3 Si 2 - U ss ingot have been cam e d out to obtain free uranium particle size improvement which is required to enhance the U-Al inter-diffusion reaction in the fuel plate meat. . Heat treatment experiments upon fuel plates containing dispersion of U 3 Si 2 - U ss in Al matrix have also been carried out to study the effect of temperature and treatment duration on the extent of inter-diffusion reaction between free uranium particle and aluminium matrix in the fuel plate meat. Both the experiments indicate that a drastic size improvement has occurred with the U 3 Si 2 as well as free uranium particles upon heat treatment at controlled temperature between the U 3 Si 2 peritectic and peritectoid temperatures and that the inter-diffusion reaction between free uranium and Al matrix occurs quite significantly at temperatures higher than that ordinarily used in the fabrication procedure

Ultrahigh flux reactor design probing the limits of plate fuel technology

International Nuclear Information System (INIS)

Lake, J.A.; Parsons, D.K.; Liebenthal, J.L.; Ryskamp, J.M.; Fillmore, G.N.; Deboisblanc, D.R.

1986-01-01

The need for a new steady-state thermal neutron source of unprecedented intensity has been the subject of numerous national meetings and discussions. The National Research Council Committee on Major Facilities for Materials Research recently issued a high priority recommendation that site-independent design studies for such a facility begin immediately. The high intensity neutron source is projected to open new frontiers in the use of neutrons as a probe in various aspects of materials and biological research and fundamental physics. The challenge put forth by the research community is to produce a source with a tenfold increase in intensity over any currently operating or planned facility and, therefore, to thrust the thermal neutron flux intensity into the 10 16 n/(cm 2 s) range. The purpose of the recent Idaho National Engineering Laboratory (INEL) activities in this area has been to identify and examine the limitations and the capabilities of the historically well-characterized plate-fuel technology to achieve the required performance levels in a user-friendly environment. Workbench design concepts were identified, upon which constraints and performance limitations could be evaluated and parametric trade-off analyses and preliminary design optimization studies could be performed. Although considerable optimization remains to be performed and a large number of cost/benefit trade-offs exist, it appears that a reactor core with innovative geometry, constructed of plate-type fuel elements, can achieve the 10 16 n/(cm 2 s) goal thermal flux level in a large external volume which has the quality and accessibility for beam research. (orig.)

Evaluation of Corrosion of the Dummy ''EE'' Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

International Nuclear Information System (INIS)

Brower, Jeffrey Owen; Glazoff, Michael Vasily; Eiden, Thomas John; Rezvoi, Aleksey Victor

2016-01-01

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and ''horseshoeing'' defects were readily observable on the surface of the several YA-type fuel elements (these are ''dummy'' plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth ''S'' curve, was represented by a series temperature rise ''humps,'' which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel elements. In the case

High enrichment to low enrichment core's conversion. Technical securities

International Nuclear Information System (INIS)

Abbate, P.; Madariaga, M.R.

1990-01-01

This work presents the fulfillment of the technical securities subscribed by INVAP S.E. for the conversion of a high enriched uranium core. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. These are neutronic and thermohydraulic securities. (Author) [es

Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

Energy Technology Data Exchange (ETDEWEB)

Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Glazoff, Michael V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Eiden, Thomas J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Rezvoi, Aleksey V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor

2016-08-01

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed

Improving 6061-Al Grain Growth and Penetration across HIP-Bonded Clad Interfaces in Monolithic Fuel Plates: Initial Studies

Energy Technology Data Exchange (ETDEWEB)

Hackenberg, Robert E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCabe, Rodney J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Randall L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trujillo, R. Ralph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lienert, Thomas J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forsyth, Robert T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harada, Kiichi L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2013-05-06

Grain penetration across aluminum-aluminum cladding interfaces in research reactor fuel plates is desirable and was obtained by a legacy roll-bonding process, which attained 20-80% grain penetration. Significant grain penetration in monolithic fuel plates produced by Hot Isostatic Press (HIP) fabrication processing is equally desirable but has yet to be attained. The goal of this study was to modify the 6061-Al in such a way as to promote a much greater extent of crossinterface grain penetration in monolithic fuel plates fabricated by the HIP process. This study documents the outcomes of several strategies attempted to attain this goal. The grain response was characterized using light optical microscopy (LOM) electron backscatter diffraction (EBSD) as a function of these prospective process modifications done to the aluminum prior to the HIP cycle. The strategies included (1) adding macroscopic gaps in the sandwiches to enhance Al flow, (2) adding engineering asperities to enhance Al flow, (3) adding stored energy (cold work), and (4) alternative cleaning and coating. Additionally, two aqueous cleaning methods were compared as baseline control conditions. The results of the preliminary scoping studies in all the categories are presented. In general, none of these approaches were able to obtain >10% grain penetration. Recommended future work includes further development of macroscopic grooving, transferred-arc cleaning, and combinations of these with one another and with other processes.

Coil-springs used as mechanical filter. Modification of the bottom tie plate of a fuel assembly

International Nuclear Information System (INIS)

Nylund, O.

1993-01-01

Describes an improved design of the bottom tie plate of a fuel assembly. The improvement of the design is an arrangement of horizontal channels holding coil-springs and crossing the vertical channels for the cooling water. The coil-springs work as strainers for the cooling water

Analysis of the production of U3O8 powder for low enrichment fuel plates

International Nuclear Information System (INIS)

Boero, N.L.; Celora, J.; Parodi, C.A.; Ponieman, G.; Kellner, M.; Marajofsky, A.

1987-01-01

Description is made of the processes used in the production of U 3 O 8 powder for low enrichment plates for fuel elements for Research Reactors. The analysis of the efficiency of each batch is foccused on the relationship between milling and sieving times and the morphology of the product in each production step. (Author)

Nuclear fuel storage

International Nuclear Information System (INIS)

Bevilacqua, F.

1981-01-01

A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

Irradiation performance of uranium-molybdenum alloy dispersion fuels

International Nuclear Information System (INIS)

Almeida, Cirila Tacconi de

2005-01-01

The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

AAA-ATPase NVL2 acts on MTR4-exosome complex to dissociate the nucleolar protein WDR74

Energy Technology Data Exchange (ETDEWEB)

Hiraishi, Nobuhiro; Ishida, Yo-ichi; Nagahama, Masami, E-mail: nagahama@my-pharm.ac.jp

2015-11-20

Nuclear VCP-like 2 (NVL2) is a chaperone-like nucleolar ATPase of the AAA (ATPase associated with diverse cellular activities) family, which exhibits a high level of amino acid sequence similarity with the cytosolic AAA-ATPase VCP/p97. These proteins generally act on macromolecular complexes to stimulate energy-dependent release of their constituents. We previously showed that NVL2 interacts with RNA processing/degradation machinery containing an RNA helicase MTR4/DOB1 and an exonuclease complex, nuclear exosome, and involved in the biogenesis of 60S ribosomal subunits. These observations implicate NVL2 as a remodeling factor for the MTR4-exosome complex during the maturation of pre-ribosomal particles. Here, we used a proteomic screen and identified a WD repeat-containing protein 74 (WDR74) as a factor that specifically dissociates from this complex depending on the ATPase activity of NVL2. WDR74 shows weak amino acid sequence similarity with the yeast ribosome biogenesis protein Nsa1 and is co-localized with NVL2 in the nucleolus. Knockdown of WDR74 decreases 60S ribosome levels. Taken together, our results suggest that WDR74 is a novel regulatory protein of the MTR4-exsosome complex whose interaction is regulated by NVL2 and is involved in ribosome biogenesis. - Highlights: • WDR74 accumulates in MTR4-exosome complex upon expression of dominant-negative NVL2. • WDR74 is co-localized with NVL2 in the nucleolus. • WDR74, along with NVL2, is involved in the synthesis of 60S ribosomal subunits.

AAA-ATPase NVL2 acts on MTR4-exosome complex to dissociate the nucleolar protein WDR74

International Nuclear Information System (INIS)

Hiraishi, Nobuhiro; Ishida, Yo-ichi; Nagahama, Masami

2015-01-01

Nuclear VCP-like 2 (NVL2) is a chaperone-like nucleolar ATPase of the AAA (ATPase associated with diverse cellular activities) family, which exhibits a high level of amino acid sequence similarity with the cytosolic AAA-ATPase VCP/p97. These proteins generally act on macromolecular complexes to stimulate energy-dependent release of their constituents. We previously showed that NVL2 interacts with RNA processing/degradation machinery containing an RNA helicase MTR4/DOB1 and an exonuclease complex, nuclear exosome, and involved in the biogenesis of 60S ribosomal subunits. These observations implicate NVL2 as a remodeling factor for the MTR4-exosome complex during the maturation of pre-ribosomal particles. Here, we used a proteomic screen and identified a WD repeat-containing protein 74 (WDR74) as a factor that specifically dissociates from this complex depending on the ATPase activity of NVL2. WDR74 shows weak amino acid sequence similarity with the yeast ribosome biogenesis protein Nsa1 and is co-localized with NVL2 in the nucleolus. Knockdown of WDR74 decreases 60S ribosome levels. Taken together, our results suggest that WDR74 is a novel regulatory protein of the MTR4-exsosome complex whose interaction is regulated by NVL2 and is involved in ribosome biogenesis. - Highlights: • WDR74 accumulates in MTR4-exosome complex upon expression of dominant-negative NVL2. • WDR74 is co-localized with NVL2 in the nucleolus. • WDR74, along with NVL2, is involved in the synthesis of 60S ribosomal subunits.

Sharing the load: Mex67-Mtr2 cofunctions with Los1 in primary tRNA nuclear export.

Science.gov (United States)

Chatterjee, Kunal; Majumder, Shubhra; Wan, Yao; Shah, Vijay; Wu, Jingyan; Huang, Hsiao-Yun; Hopper, Anita K

2017-11-01

Eukaryotic transfer RNAs (tRNAs) are exported from the nucleus, their site of synthesis, to the cytoplasm, their site of function for protein synthesis. The evolutionarily conserved β-importin family member Los1 (Exportin-t) has been the only exporter known to execute nuclear export of newly transcribed intron-containing pre-tRNAs. Interestingly, LOS1 is unessential in all tested organisms. As tRNA nuclear export is essential, we previously interrogated the budding yeast proteome to identify candidates that function in tRNA nuclear export. Here, we provide molecular, genetic, cytological, and biochemical evidence that the Mex67-Mtr2 (TAP-p15) heterodimer, best characterized for its essential role in mRNA nuclear export, cofunctions with Los1 in tRNA nuclear export. Inactivation of Mex67 or Mtr2 leads to rapid accumulation of end-matured unspliced tRNAs in the nucleus. Remarkably, merely fivefold overexpression of Mex67-Mtr2 can substitute for Los1 in los1 Δ cells. Moreover, in vivo coimmunoprecipitation assays with tagged Mex67 document that the Mex67 binds tRNAs. Our data also show that tRNA exporters surprisingly exhibit differential tRNA substrate preferences. The existence of multiple tRNA exporters, each with different tRNA preferences, may indicate that the proteome can be regulated by tRNA nuclear export. Thus, our data show that Mex67-Mtr2 functions in primary nuclear export for a subset of yeast tRNAs. © 2017 Chatterjee et al.; Published by Cold Spring Harbor Laboratory Press.

In-vivo identification of direct electron transfer from Shewanella oneidensis MR-1 to electrodes via outer-membrane OmcA-MtrCAB protein complexes

Energy Technology Data Exchange (ETDEWEB)

Okamoto, Akihiro [Department of Applied Chemistry, School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Nakamura, Ryuhei, E-mail: nakamura@light.t.u-tokyo.ac.jp [Department of Applied Chemistry, School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Hashimoto, Kazuhito, E-mail: hashimoto@light.t.u-tokyo.ac.jp [Department of Applied Chemistry, School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); ERATO/JST, HASHIMOTO Light Energy Conversion Project (Japan)

2011-06-30

Graphical abstract: . Display Omitted Highlights: > Monolayer biofilm of Shewanella cells was prepared on an ITO electrode. > Extracellular electron transfer (EET) process was examined with series of mutants. > Direct ET was confirmed with outer-membrane-bound OmcA-MtrCAB complex. > The EET process was not prominently influenced by capsular polysaccharide. - Abstract: The direct electron-transfer (DET) property of Shewanella bacteria has not been resolved in detail due to the complexity of in vivo electrochemistry in whole-cell systems. Here, we report the in vivo assignment of the redox signal indicative of the DET property in biofilms of Shewanella oneidensis MR-1 by cyclic voltammetry (CV) with a series of mutants and a chemical marking technique. The CV measurements of monolayer biofilms formed by deletion mutants of c-type cytochromes ({Delta}mtrA, {Delta}mtrB, {Delta}mtrC/{Delta}omcA, and {Delta}cymA), and pilin ({Delta}pilD), capsular polysaccharide ({Delta}SO3177) and menaquinone ({Delta}menD) biosynthetic proteins demonstrated that the electrochemical redox signal with a midpoint potential at 50 mV (vs. SHE) was due to an outer-membrane-bound OmcA-MtrCAB protein complex of decaheme cytochromes, and did not involve either inner-membrane-bound CymA protein or secreted menaquinone. Using the specific binding affinity of nitric monoxide for the heme groups of c-type cytochromes, we further confirmed this conclusion. The heterogeneous standard rate constant for the DET process was estimated to be 300 {+-} 10 s{sup -1}, which was two orders of magnitude higher than that previously reported for the electron shuttling process via riboflavin. Experiments using a mutant unable to produce capsular polysaccharide ({Delta}SO3177) revealed that the DET property of the OmcA-MtrCAB complex was not influenced by insulating and hydrophilic extracellular polysaccharide. Accordingly, under physiological conditions, S. oneidensis MR-1 utilizes a high density of outer

Reprocessing of research reactor fuel the Dounreay option

Energy Technology Data Exchange (ETDEWEB)

Cartwright, P.

1997-08-01

Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.