WorldWideScience
1

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel  

Energy Technology Data Exchange (ETDEWEB)

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.

A K Wertsching

2012-09-01

2

Fabrication results of gamma uranium-molybdenum alloys fuels  

International Nuclear Information System (INIS)

This paper describes the results on the development of the technology of the fabrication of the gamma uranium molybdenum alloys in IPEN-CNEN-SP, and presents some of their more recent experimental results. The importance of this class of fuels relies on the fact that they are the fuels considered to be loaded in the first Brazilian Multipurpose Reactor, RMB, stated as one of the tasks in the Nuclear Brazilian Plan, PNB. The study of ? UMo fuels started with their preparation by the arc and induction melting technique, followed by thermal treatment to the obtention of a better degree of homogenization, under argon atmosphere at 1000 deg C. Additions of Mo varied from 5 to 10% weight. Samples of both classes of fuels were characterized mainly by X-ray diffraction, density, SEM and optical microscopy with image analysis, The main results of the alloy's production and an emphasis of the use of XRD data in the gamma-UMo powder obtention process are presented and emphasized here. The results enabled us to study future methodologies to avoid most of the problems encountered in the recent technological approach to the fabrication of the alloys of UMo, which will lead to the production of materials with best efficiency and quality. (author)

3

Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2  

International Nuclear Information System (INIS)

This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ?40% and ?70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

4

Irradiation performance of uranium-molybdenum alloy dispersion fuels  

International Nuclear Information System (INIS)

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/cm3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

5

CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL  

Energy Technology Data Exchange (ETDEWEB)

The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for establishing preconceptual fabrication facility designs.

Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

2008-02-01

6

Update on uranium-molybdenum fuel foil fabrication development activities at the Y-12 National Security Complex in 2007  

International Nuclear Information System (INIS)

In support of the RERTR Program, efforts are underway at Y-12 to develop and validate a production oriented, monolithic uranium molybdenum (U-Mo) foil fabrication process adaptable for potential implementation in a manufacturing environment. These efforts include providing full-scale prototype depleted and enriched U-Mo foils in support of fuel qualification testing. The work has three areas of focus; develop and demonstrate a feasible foil fabrication process utilizing depleted uranium-molybdenum (DU-Mo) source material, transition these production techniques to enriched uranium (EU-Mo) source material, and evaluate full-scale implementation of the developed production techniques. In 2006, Y-12 demonstrated successful fabrication of full-size DU-10Mo foils. In 2007, Y-12 activities were expanded to include continued DU-Mo foil fabrication with a focus on process refinement, source material impurity effects (specifically carbon), and the feasibility of physical vapor deposition (PVD) on DU-10Mo mini-foils. FY2007 activities also included a transition to EU-Mo and fabrication of full-size enriched foils. The purpose of this report is to update the RERTR audience on Y-12 efforts in 2007 that support the overall RERTR Program goals. (author)

7

Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop  

Energy Technology Data Exchange (ETDEWEB)

Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-Mo alloy fuel at a workshop held at Argonne National Laboratory on January 17-18, 2000. Consensus was reached that the qualification plans of the U.S. RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper. (author)

Snelgrove, J.L. [Argonne National Laboratory, Argonne (United States); Languille, A. [CEA Cadarache, F-13108 Saint Paul lez Durance (France)

2000-07-01

8

Qualification of uranium-molybdenum alloy fuel -- conclusions of an international workshop  

Energy Technology Data Exchange (ETDEWEB)

Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-MO alloy fuel at a workshop held at Argonne National Laboratory on January 17--18, 2000. Consensus was reached that the qualification plans of the US RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper.

Snelgrove, J. L.; Languilee, A.

2000-02-14

9

Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures  

International Nuclear Information System (INIS)

This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and pre-treatment to stabilize the gamma structure. The addition of a bit low ternary excess and formation of an intergranular phase, the increase in stability, it was demonstrated that there is not a damage in the formation of their powders.(author)

10

Post-irradiation examination of uranium-molybdenum dispersion fuel irradiated to high burn-up in NRU  

International Nuclear Information System (INIS)

UMo dispersion fuels are promising candidates for research and test reactors. Mini-elements containing U7Mo and U10Mo (7 and 10 wt% Mo in U alloy) fuel particles dispersed in aluminium have been fabricated with a nominal loading of 4.5 gU/cm3. In order to compare the performance of the different UMo alloys, the mini-elements were irradiated adjacent to each other under nominally identical conditions in the National Research Universal (NRU) reactor. Maximum element linear ratings up to 100 kW/m and discharge burnups up to 80 atom% 235U were achieved. The experiment was conducted in phases such that adjacent pairs of mini-elements could be removed for post-irradiation examinations (PIE) after 20, 40, 60 and 80 atom% 235U burnup. PIE included underwater inspections, visual examinations and photography in the hot cells, gamma spectroscopy, dimensional measurements, immersion density measurements, metallography, and chemical burnup analysis. The results from the high burnup fuels are presented in this paper. The assessments compare the microstructural changes, porosity formation and fuel swelling in the two UMo dispersion fuels. The results indicate that U7Mo fuel is less stable that U10 Mo fuel under the conditions tested in NRU. (author)

11

Spectrographic analysis of uranium-molybdenum alloys  

International Nuclear Information System (INIS)

A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO3. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO3. (Author) 5 refs

12

Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas  

Energy Technology Data Exchange (ETDEWEB)

This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and pre-treatment to stabilize the gamma structure. The addition of a bit low ternary excess and formation of an intergranular phase, the increase in stability, it was demonstrated that there is not a damage in the formation of their powders.(author)

Oliveira, Fabio Branco Vaz de

2008-07-01

13

Qualification of Uranium-Molybdenum Alloys for Research Reactor Community  

Energy Technology Data Exchange (ETDEWEB)

Uranium-molybdenum (U-Mo) alloys are being produced to refuel international research reactors - replacing current highly-enriched uranium fuel assemblies. Over the past two years, Y-12 Analytical Chemistry has been the primary qualification laboratory for current U-Mo materials development in the U.S. During this time, multiple analytical techniques have been explored to obtain complete and accurate characterization of U-Mo materials. For the chemical characterization of U-Mo materials, three primary techniques have been utilized: (i) thermal ionization mass spectrometry (TIMS) for uranium content and isotopic analyses, (ii) a combination of inductively-coupled plasma (ICP) techniques for determination of molybdenum content and trace elemental concentrations and (iii) combustion analyses for trace elemental analyses. Determination of uranium content, uranium isotopic composition and elemental impurities by combustion analyses (H, C, O, N) required only minimal changes to existing analytical methodology for uranium metal analyses. However, spectral interferences (both isobaric and optical) due to high molybdenum content presented significant challenges to the use of ICP instrumentation. While providing a brief description of methods for determination of uranium content and H, C, O and N content, this manuscript concentrates on the challenges faced in applying ICP techniques to qualification of U-Mo fuels. Multiple ICP techniques were explored to determine the effectiveness (e.g., accuracy, precision, speed of analysis, etc.) for determining both molybdenum content and trace elemental impurity concentrations: high-resolution inductively-coupled plasma mass spectrometry (HR-ICPMS), inductively- coupled plasma quadrupole mass spectrometry (ICP-QMS) and inductively-coupled plasma optical emission spectroscopy (ICP-OES). The merits and limitations of these techniques for qualification of U-Mo alloys are presented, to include the limits of quantitation and uncertainties of measurements regarding the most efficient methods for qualifying the U-Mo alloys. (author)

Schaaff, T.G.; Belt, V.F.; Likens, A.M.; Joyce, K.E.; Barry, J.F. [Analytical Chemistry Organization, Y-12 National Security Complex, Oak Ridge, Tennessee 37831 (United States)

2011-07-01

14

Structural control and zonality of uranium-molybdenum ores of Kurdai Deposit (Southern Kazakhstan)  

International Nuclear Information System (INIS)

General characteristics of the Kurdai uranium-molybdenum deposit (Souther Kazakhstan) is given, internal structure of the Kurdai fracture is considered. Structural conditions of the mineralization localization are discussed. Attention is paid to the vertical structural and ore zonation

15

Powder formation of ? uranium-molybdenum alloys via hydration-dehydration  

International Nuclear Information System (INIS)

Gamma uranium-molybdenum alloys has been considered as fuel phase in plate type fuel elements for MTR reactors, mainly due to their acceptable performance under irradiation and metallurgical processing. To its use as a dispersion phase in aluminum matrix, a necessary step is the conversion of the as cast structure into powder, and one of the techniques considered at IPEN / CNEN - Brazil is HDH (hydration-dehydration). The alloys were produced by the induction melting technique, and samples were obtained from the alloys for the thermal treatments, under constant flow of hydrogen, for temperatures varying from 400 deg C to 600 deg C and times from 1 to 4 hours, followed by dehydration. A preliminary characterization of the powders was made and the curves of mass variation versus time were obtained and related to the powder characteristics. This paper describes the first results on the development of the technology to the powder formation of the (5 to 10) % weight molybdenum ?-UMo alloys, and discusses some of its aspects, mainly those related to the ? ? ? equilibrium data. (author)

16

Study and comparison of analytical methods for dosing molybdenum in uranium-molybdenum alloys  

International Nuclear Information System (INIS)

Methods to determine molybdenum in uranium-molybdenum alloys are developed by various technic: molecular absorption spectrophotometry, emission spectroscopy, X ray fluorescence, atomic absorption spectrophotometry. After a comparison on samples in which molybdenum content lies between 1 and 10 per cent by weight, one concludes in the interest of some of the exposed methods for routine analysis. (author)

17

Environmental impact study report, Ben Lomond uranium-molybdenum project, Northern Queensland  

International Nuclear Information System (INIS)

A significant uranium-molybdenum mineralisation has been discovered in Northern Queensland, west of Townsville. Granting of a mining lease is subject to the compilation and acceptance of an environmental impact study report. The report describes the proposed mining and milling project, the existing environment and the impact of the proposal on the environment. Two main environmental safeguards incorporated into the project are a comprehensive water management scheme and a progressive site rehabilitation

18

Investigation of the uranium-molybdenum diffusion in body centered ? solid solutions  

International Nuclear Information System (INIS)

The body centered ? phase uranium-molybdenum intermetallic diffusion has been studied by different technical methods: micrography, electronic microanalyser, microhardness. The values of several numbers of penetration coefficients are given, and their physical significations has been discussed. The diffusion coefficients, the frequency factor and activation energies has been determined for each concentration. After determination of the Kirkendall effect in this system, we calculated the intrinsic diffusion coefficient of uranium and molybdenum. (author)

19

Surface engineering of low enriched uranium–molybdenum  

International Nuclear Information System (INIS)

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)–Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (?50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations

20

Surface engineering of low enriched uranium–molybdenum  

Energy Technology Data Exchange (ETDEWEB)

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% {sup 235}U. The root cause of the failures is clearly related directly to the formation of the U(Mo)–Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm{sup 2} until a maximum local burn-up of approximately 70% {sup 235}U (?50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

Leenaers, A., E-mail: aleenaer@sckcen.be [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van den Berghe, S. [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)

2013-09-15

21

Surface engineering of low enriched uranium-molybdenum  

Science.gov (United States)

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

Leenaers, A.; Van den Berghe, S.; Detavernier, C.

2013-09-01

22

Study of ??? and ??? transformations in low content uranium-molybdenum alloys  

International Nuclear Information System (INIS)

Plots of the TTT graphs corresponding to the ??? and ??? transformations have been established for two uranium-molybdenum alloys containing 1.1 and 2.1 per cent by weight of molybdenum. These TTT curves have the characteristics of the transformations due to germination and growth, however: - they are flat and almost linear in the ??? zone whereas they have the usual C shape in the ??? zone; - the incubation periods for the ??? transformation lengthen rapidly with increasing temperature of transformation; those for the ??? transformation are very short. - the curves for the ??? transformations exhibit no continuous solution at a temperature of 645 C. The beginning of the transformations in the three zones (? + ?), (? + ?), (? + ?') can be represented by a single C-shaped; - the distance between the curves at the beginning and at the end of the ??? transformation gets progressively smaller as the transformation temperature increases. A delay in the transformation has been observed when the alloys have undergone a homogenizing treatment (950 C) which leads to a large-grain ? structure. (author)

23

The Problem of Storing Fission Products Arising from the Processing of Irradiated Uranium-Molybdenum Alloys  

International Nuclear Information System (INIS)

Uranium-molybdenum alloys are of value thanks to their in-pile behaviour but serious disadvantages arise in connection with the storing of fission products resulting from the processing of these alloys. Because of the insolubility of molybdenum it is impossible to concentrate a solution of fission products by evaporation, and for this reason we have directed our efforts towards the solubilization of molybdenum through the addition of reagents such as iron or phosphoric ions. In this way one can obtain final solutions of 60 g/l Mo with Fe 100 g/l Mo with PO4H3. The volumes to be stored are still considerable (especially with Fe) and the possibility of nitrate calcination in a fluidized bed was considered. The reaction takes place at about 400°C. The behaviour of the ruthenium and the friability of the calcined solid (formation of considerable amounts of fine material) have led us to abandon this process in favour of the preparation of phosphate glasses. (author)

24

Application of comprehensive geophysical and geochemical survey method in the exploration of uranium-molybdenum deposit 460  

International Nuclear Information System (INIS)

This paper summarized the application effect of geophysical and geochemical survey method in uranium-molybdenum deposit 460. It stress on illustrating the effects of induced current middle gradient, high precision magnetic survey and gravity survey method to identify the distribution features of fracture, volcano structure and sub-rhyolite porphyry. Through verifying the mineralization caused anomaly which measured by activated charcoal, gamma, uranium content and secondary halo in soil with borehole, good prospecting result was achieved. Based on the above application effect, the paper presented some helpful prospection method combination. (authors)

25

A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys  

International Nuclear Information System (INIS)

Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the ? and ? solid solutions are described. These processes depend upon molybdenum concentration. Out of the ? solid solution phase appears an eutectoid decomposition of ? to (? + ?) or the formation of a martensitic phase ?''. The ? solid solution shows a decomposition of ? to (? + ?) or (? + ?'), or a formation of martensitic phases a' or a'b. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains (? + ?) and (? + ?), (? + ?) and ?, (? + ?) and ?, have been determined. (author)

26

Obtention of uranium-molybdenum alloy ingots technique to avoid carbon contamination  

Energy Technology Data Exchange (ETDEWEB)

The replacement of high enriched uranium (U{sup 235} > 85 wt%) by low enriched uranium (U{sup 235} < 20wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Among the several uranium alloys investigated since then, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloy is being performed at the Nuclear Technology Development Centre (CDTN) and also at IPEN. The carbon contamination of the alloy is one of the great concerns during the melting process. It was observed that U-Mo alloy is more critical considering carbon contamination when using graphite crucibles. Alternative melting technique was implemented at CDTN in order to avoid carbon contamination from graphite crucible using Yttria stabilized ZrO{sub 2} crucibles. Ingots with low carbon content and good internal quality were obtained. (author)

Pedrosa, Tercio A.; Paula, Joao Bosco de; Reis, Sergio C.; Brina, Jose Giovanni M.; Faeda, Kelly Cristina M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

27

Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization  

Energy Technology Data Exchange (ETDEWEB)

The replacement of high enriched uranium (U-{sup 235} > 85 wt%) by low enriched uranium (U-{sup 235} < 20 wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Several uranium alloys that fill this requirement has been investigated since then. Among these alloys, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloys is being performed at the Nuclear Technology Development Centre (CDTN) and also at the Institute of Energetic and Nuclear Research - IPEN. U-{sup 10}Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U{sub 2}Mo compound and alpha-phase, in the heat treated condition. (author)

Pedrosa, Tercio A.; Braga, Daniel M.; Paula, Joao Bosco de; Brina, Jose Giovanni M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: bragadm@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

28

Investigation of the uranium-molybdenum diffusion in body centered {gamma} solid solutions; Etude de la diffusion uranium-molybdene dans la solution solide {gamma} cubique centree  

Energy Technology Data Exchange (ETDEWEB)

The body centered {gamma} phase uranium-molybdenum intermetallic diffusion has been studied by different technical methods: micrography, electronic microanalyser, microhardness. The values of several numbers of penetration coefficients are given, and their physical significations has been discussed. The diffusion coefficients, the frequency factor and activation energies has been determined for each concentration. After determination of the Kirkendall effect in this system, we calculated the intrinsic diffusion coefficient of uranium and molybdenum. (author) [French] La dilution intermetallique uranium-molybdene, en phase {gamma} cubique centree, a ete etudiee au moyen de differentes techniques: micrographie, microsonde electronique, microdurete. Les valeurs d'un certain nombre de coefficients de penetration sont donnees et leur signification physique discutee. Les coefficients de diffusion, les facteurs de frequence et les energies d'activation ont ete determines pour chaque concentration. Apres avoir mis en evidence un effet Kirkendall dans ce systeme, on a calcule les coefficients de diffusion intrinseques de l'uranium et du molybdene. (auteur)

Adda, Y.; Mairy, C.; Bouchet, P. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Philibert, J. [IRSID, 78 - Saint-Germain-en-Laye (France)

1958-07-01

29

UPDATE ON MONOLITHIC FUEL FABRICATION METHODS  

Energy Technology Data Exchange (ETDEWEB)

Efforts to develop a viable monolithic research reactor fuel plate have continued at Idaho National Laboratory. These efforts have concentrated on both fabrication process refinement and scale-up to produce full sized fuel plates. Progress at INL has led to fabrication of hot isostatic pressed uranium-molybdenum bearing monolithic fuel plates. These miniplates are part of the RERTR-8 miniplate irradiation test. Further progress has also been made on friction stir weld processing which has been used to fabricate full size fuel plates which will be irradiated in the ATR and OSIRIS reactors.

C. R. Clark; J. F. Jue; G. A. Moore; N. P. Hallinan; B. H. Park; D. E. Burkes

2006-10-01

30

Update on US High Density Fuel Fabrication Development  

Energy Technology Data Exchange (ETDEWEB)

Second generation uranium molybdenum fuel has shown excellent in-reactor irradiation performance. This metallic fuel type is capable of being fabricated at much higher loadings than any presently used research reactor fuel. Due to the broad range of fuel types this alloy system encompasses—fuel powder to monolithic foil and binary fuel systems to multiple element additions—significant amounts of research and development have been conducted on the fabrication of these fuels. This paper presents an update of the US RERTR effort to develop fabrication techniques and the fabrication methods used for the RERTR-9A miniplate test.

C.R. Clark; G.A. Moore; J.F. Jue; B.H. Park; N.P. Hallinan; D.M. Wachs; D.E. Burkes

2007-03-01

31

Update on fuel fabrication development and testing at Argonne National Laboratory  

International Nuclear Information System (INIS)

In its effort to develop research reactor fuel with a high fissile loading, Argonne National Laboratory has continued its advanced fuel development efforts. Monolithic fuel, where the fuel is in the form of a single fuel foil, is being developed as the ultimate in fuel loading capacity. Work has been done on different monolithic fabrication methods that have resulted in process refinements. Effort is also underway to develop a uranium-molybdenum dispersion fuel plate that will be resistant to the irradiation shortcomings noted in previous tests. Alloying additions to the aluminum matrix are being investigated. These fuels are being fabricated for use in irradiation experiments scheduled for insertion in 2005. (author)

32

Development and validation of capabilities to measure thermal properties of layered monolithic U-Mo alloy plate-type fuel  

Energy Technology Data Exchange (ETDEWEB)

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of thermal conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify and validate the functionality of equipment methods installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, procedures to operate the equipment, and models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a zirconium diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-19

33

Complex plasmochemical processing of solid fuel  

Directory of Open Access Journals (Sweden)

Full Text Available Technology of complex plasmaochemical processing of solid fuel by Ecibastuz bituminous and Turgay brown coals is presented. Thermodynamic and experimental study of the technology was fulfilled. Use of this technology allows producing of synthesis gas from organic mass of coal and valuable components (technical silicon, ferrosilicon, aluminum and silicon carbide and microelements of rare metals: uranium, molybdenum, vanadium etc. from mineral mass of coal. Produced a high-calorific synthesis gas can be used for methanol synthesis, as high-grade reducing gas instead of coke, as well as energy gas in thermal power plants.

Vladimir Messerle

2012-12-01

34

Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens  

Energy Technology Data Exchange (ETDEWEB)

The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

2012-10-01

35

A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys; Etude des processus des transformations dans les alliages uranium-molybdene de teneur 0,5 a 4% en poids de molybdene  

Energy Technology Data Exchange (ETDEWEB)

Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the {beta} and {gamma} solid solutions are described. These processes depend upon molybdenum concentration. Out of the {beta} solid solution phase appears an eutectoid decomposition of {beta} to ({alpha} + {gamma}) or the formation of a martensitic phase {alpha}''. The {gamma} solid solution shows a decomposition of {gamma} to ({alpha} + {gamma}) or ({alpha} + {gamma}'), or a formation of martensitic phases a' or a'{sub b}. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains ({alpha} + {gamma}) and ({beta} + {gamma}), ({beta} + {gamma}) and {gamma}, ({beta} + {gamma}) and {beta}, have been determined. (author) [French] Les processus des transformations isothermes, et au cours de refroidissements continus ont ete etablis sur les alliages uranium-molybdene de 0,5 a 4 % en poids de Mo. Ceci a permis de mettre en evidence les processus des transformations de solutions solides {beta} et {gamma}, differents suivant la teneur en molybdene de l'alliage. Dans le premier cas il y a decomposition eutectoide de {beta} en ({alpha} + {gamma}) ou formations d'une phase martensitique {alpha}''. Dans le second cas il y a decomposition de {gamma} soit en ({alpha} + {gamma}) soit en ({alpha} + {gamma}') suivant la temperature, ou bien formation des phases martensitiques {alpha}' ou {alpha}'{sub b}. Le diagramme d'equilibre, uranium-molybdene est sujet a de nombreuses controverses, en particulier dans la zone des faibles concentrations. Les limites entre les domaines ({alpha} + {gamma}) et ({beta} + {gamma}), ({beta} + {gamma}) et {gamma}, ({beta} + {gamma}) et {beta}, ont ete determinees. (auteur)

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

1959-06-15

36

A new fuel for research reactors  

International Nuclear Information System (INIS)

The Replacement Research Reactor (RRR) to be constructed at Lucas Heights will use fuel containing low enriched uranium (LEU), 235U, whereas its predecessor HIFAR operates with fuel fabricated from high-enriched uranium (HEU). The fuel will be based on uranium silicide (U3Si2) with a density of 4.8 g U/cm3. This fuel has been qualified and in use in 20 research reactors worldwide for over 12 years A brief description is given of the metallurgy, behaviour under irradiation, and fabrication methods, all of which are well-understood Progress on development of new, higher density LEU fuel based on uranium molybdenum alloys is also described and the implications for the RRR discussed briefly

37

Contribution to the study of remedy solutions to uranium(molybdenum)/aluminium interactions: role of silicon addition to aluminium, study of coupled effects  

International Nuclear Information System (INIS)

In the project development and qualification program of a nuclear fuel with Low Enriched Uranium for Materials Testing Reactors, the dispersed U(Mo)/Al fuel is being developed due to its excellent stability during irradiation. However, in pile experiments showed that depending on the irradiation conditions (e.g. high burnup or high heat flux), an extensive interaction occurs between the fissile element U(Mo) and the Al based matrix resulting in swelling, which could eventually lead to a fuel plate failure. Among the ways to improve the behavior of the dispersed U(Mo) fuel, the solution now seen as the reference remedy by the entire scientific community is the addition of silicon into the aluminum matrix. In order to provide some understanding and optimizing the solution 'Si additions into Al matrix' under neutron irradiation, an out of pile study is performed on (i) the interaction mechanisms involved in the U(Mo)/Al (Si) system and (ii) the impact of the Si additions into the Al matrix on alternative solutions to the U(Mo)/Al interactions, namely the modification of the ?-U(Mo) fissile compound by adding a third element and/or modifying the interface between the ?-U(Mo) fissile compound and the matrix. This document provides a mechanistic description of the U(7Mo)/Al(Si) interaction for a range of Si content in Al between 2 and 10 wt.%, based on the multi-scale characterization of diffusion couples. The location of the Mo and its role in the reaction mechanisms are demonstrated. The influence of elements X = Y, Cu, Zr, Ti, Cr, on the U (Mo)/Al and U (Mo)/Al (Si) interactions mechanisms was then studied. It is shown that adding a third element to the U(Mo) alloy acts on the second order on diffusion kinetics and (micro)structure of the interaction layer compared to the addition of Si into Al. Finally, an alumina coating which shows a potential interest to improve the performance of the fuel has been developed. (author)

38

Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites  

Energy Technology Data Exchange (ETDEWEB)

The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

Burkes, Douglas; Casella, Amanda J.; Gardner, Levi D.; Casella, Andrew M.; Huber, Tanja K.; Breitkreutz, Harald

2015-02-11

39

Measurement of fission gas release from irradiated U–Mo monolithic fuel samples  

Energy Technology Data Exchange (ETDEWEB)

The uranium–molybdenum (U–Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium–molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30–1000 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

Douglas E. Burkes; Amanda J. Casella; Andrew M. Casella; Karl N. Pool; Francine J. Rice

2015-02-01

40

Design and Testing of Prototypic Elements Containing Monolithic Fuel  

Energy Technology Data Exchange (ETDEWEB)

The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

2011-10-01

41

Research reactor fuel - an update  

International Nuclear Information System (INIS)

In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

42

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

Science.gov (United States)

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 ?m 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.

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.; King, Jeffrey C.

2014-10-01

43

Morning light cleanup and recovery operation: simulation studies of possible reactor fuels  

International Nuclear Information System (INIS)

The nuclear fuel for Cosmos 954, the orbiting Russian reactor that broke up on reentry during January of 1978, has been identified as a U--Mo alloy containing about 10 wt% molybdenum. Identification was based on a combination of simulation studies at LLL, examination of fuel debris at Whiteshell Nuclear Research Establishment (WNRE), Pinawa, Manitoba, and reactor technology knowledge. In the LLL simulation studies, mixtures of uranium, molybdenum, and UO2 were heated under conditions that simulated reentry and then examined by scanning electron microscopy, energy dispersive spectrometry, and x-ray diffraction. These studies indicated metallic behavior and suggested a U--Mo alloy. The identification was useful in assisting the Canadians in recovery, cleanup, and health/safety activities associated with the radioactive debris, which was scattered over a wide region of the Great Slave Lake

44

Modeling thermal and stress behavior of the fuel-clad interface in monolithic fuel mini-plates  

International Nuclear Information System (INIS)

A fuel development and qualification program is in process with the objective of qualifying very high density monolithic low enriched uranium-molybdenum fuel for high-performance research reactors. The monolithic fuel foil creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. 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 an unconstrained fuel plate configuration is greatly enhanced in a constrained fuel plate configuration. The sensitivities of the model and input parameters are discussed, along with some overlap of initial experimental observations using as-fabricated plate characterization and post-irradiation examination.

45

Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance  

International Nuclear Information System (INIS)

Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

46

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel  

International Nuclear Information System (INIS)

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U3O8 mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties

47

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel  

Energy Technology Data Exchange (ETDEWEB)

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

2006-02-01

48

Neutronic comparison of the nuclear fuels U3Si2/Al and U-Mo/Al  

International Nuclear Information System (INIS)

The search for materials that allow the fabrication of nuclear fuels with higher uranium densities comes from the mid 50s. Today, a high density and low enriched nuclear fuel based on ?-UMo alloys is the most promising fuel to replace the U3Si2/Al dispersion fuel used worldwide in research and material test reactors. Alloys of uranium-molybdenum are prepared with 6 to 10% Mo addition and can be manufactured as dispersion or monolithic fuels. The aim of this paper is to compare the infinite multiplication factor (K?), obtained through neutronic calculation with the code Scale 6, for aluminum coated plates reflected in all directions containing U3Si2/Al and U-Mo/Al dispersion fuels. The U3Si2/Al dispersion fuel used in the calculation has an uranium density of 4 gU/cm3 and the U-Mo-Al dispersion fuels have densities ranging from 4 to 7.52 gU/cm3 and 7 and 10% Mo addition. The results show that the K? calculated for U-Mo/Al fuels is lower than that for U3Si2/Al fuel and increases between the uranium densities of 4 and 5 gU/cm3 and decreases for higher uranium densities. (author)

49

Uranium-Molybdenum particles produced by electro-erosion  

International Nuclear Information System (INIS)

We have produced spheroidal U-Mo particles by the electro-erosion method using pure water as dielectric. The particles were characterised by optical metallography, scanning electron microscopy, energy dispersive spectrometry (EDS-EDAX) and X-ray diffraction. Spheroidal UO2 particles with a peculiar distribution size were obtained with two distribution centred at 10 and 70 ?m. The obtained particles have central inclusions of U and Mo compounds. (author)

50

Volcanogenic uranium-molybdenum deposits in Russia and China  

International Nuclear Information System (INIS)

Recent information on uranium deposits in Russia and China is presented. Most of the new information is on volcanogenic uranium deposits which, in terms of magnitude of resources, probably are the most important of all types in Russia and the second most important in China. Little or no information is available on other types of deposits in these two countries. Much of the information on China was obtained from Chinese visitors this year, and it provides our first insight into the geology of some of their more important uranium deposits. The information is far from complete as reflected by frequent omissions of location, scale, magnitude of resources, etc., in the following reviews. Those familiar with deposits in the McDermitt caldera on the Nevada-Oregon boundary and at Pena Blanca in Chihuahua, Mexico, will undoubtedlly perceive similarities in the geologic environment and ore controls. Others may also be intrigued by some aspects of similarity with deposits at Marysvale, Utah; Lake City, Colorado; Lakeview, Oregon; and elsewhere in the United States

51

Improving the performance of the performance of U-Mo fuels  

International Nuclear Information System (INIS)

Full text: Recent developments showed that uranium-molybdenum nuclear fuel particles dispersed in an aluminum matrix had misbehavior when irradiated at high neutron fluxes. The appearance of a third phase, with the presence of great porosity in the interaction zone of the Al/U-Mo interface, conditions severely the performance of this fuel. At the light of the resolution of this limitation, UMo monolithic fuel achieves a greater importance, since there is some expectation that in this bulk geometry the problem will not be present. From the simplest point of view, the addition of extra alloys to the aluminum matrix or to the nuclear fuel can be an alternative to reduce the interface growing kinetics and thereafter the appearance of the problematic third phase. The kinetics reduction would be a quantitative effect controlling chemical potentials (diffusion driving force) and barely will avoid the problem. Similar considerations can be attributed to the monolithic fuel if only quantitative solutions are proposed. In this paper are presented two drastical alternatives, from the point of view of qualitative metallurgy, for increasing the performance of U-Mo fuels. The first one is related with the coverage of the fuel particles with compound diffusion barriers to avoid the transportation of uranium and aluminum threw them. The second alternative is a monolithic fuel with zircaloy cladding where the interaction is much smaller than with aluminum. (author)ith aluminum. (author)

52

Aqueous processing of U-10Mo scrap for high performance research reactor fuel  

Science.gov (United States)

The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO3)3. HNO3 and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr2 intermetallic, explosive complex, while meeting the requirements needed for further processing.

Youker, Amanda J.; Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F.

2012-08-01

53

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

International Nuclear Information System (INIS)

UMo fuels are considered by the RERTR programme because of their higher density as compared to U3Si2. 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)

54

Neutronic comparison of the nuclear fuels U{sub 3}Si{sub 2}/Al and U-Mo/Al  

Energy Technology Data Exchange (ETDEWEB)

The search for materials that allow the fabrication of nuclear fuels with higher uranium densities comes from the mid 50s. Today, a high density and low enriched nuclear fuel based on ?-UMo alloys is the most promising fuel to replace the U{sub 3}Si{sub 2}/Al dispersion fuel used worldwide in research and material test reactors. Alloys of uranium-molybdenum are prepared with 6 to 10% Mo addition and can be manufactured as dispersion or monolithic fuels. The aim of this paper is to compare the infinite multiplication factor (K?), obtained through neutronic calculation with the code Scale 6, for aluminum coated plates reflected in all directions containing U{sub 3}Si{sub 2}/Al and U-Mo/Al dispersion fuels. The U{sub 3}Si{sub 2}/Al dispersion fuel used in the calculation has an uranium density of 4 gU/cm{sup 3} and the U-Mo-Al dispersion fuels have densities ranging from 4 to 7.52 gU/cm{sup 3} and 7 and 10% Mo addition. The results show that the K? calculated for U-Mo/Al fuels is lower than that for U{sub 3}Si{sub 2}/Al fuel and increases between the uranium densities of 4 and 5 gU/cm{sup 3} and decreases for higher uranium densities. (author)

Muniz, Rafael O.R.; Domingos, Douglas B.; Santos, Adimir dos; Silva, Antonio T. e; Joao, Thiago G.; Aredes, Vitor O., E-mail: romuniz@usp.br, E-mail: douglasborgesdomingos@gmail.com, E-mail: asantos@ipen.br, E-mail: teixeira@ipen.br, E-mail: thgarciaj@gmail.com, E-mail: vitoraredes@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

55

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011  

Energy Technology Data Exchange (ETDEWEB)

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01

56

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011  

International Nuclear Information System (INIS)

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

57

Progress in the development of very high density research and test reactor fuels  

International Nuclear Information System (INIS)

New nuclear fuels are being developed to enable many of the most important research and test reactors worldwide to convert from high enriched uranium (HEU) fuels to low enriched uranium (LEU) fuels without significant loss in performance. The last decade of work has focused on the development of uranium-molybdenum alloy (U-Mo) based fuels and is an international effort that includes the active participation of more than ten national programs. The US RERTR program, under the NNSA's Global Threat Reduction Initiative (GTRI), is in the process of developing both dispersion and monolithic U-Mo fuel designs. While the U-Mo fuel alloy has behaved extremely well under irradiation, initial testing (circa 2003) revealed that the U-Mo fuels dispersed in aluminum had an unexpected tendency toward unstable swelling (pillowing) under high-power conditions. Technical investigations were initiated worldwide at this time by the partner programs to understand this behavior as well as to develop and test remedies. The behavior was corrected by modifying the chemistry of the U-Mo/Al interfaces in both fuel designs. In the dispersion fuel design, this was accomplished by the addition of small amounts of silicon to the aluminum matrix material. Two methods are under development for the monolithic fuel design, which include the application of a thin layer of silicon or a thin zirconium based diffusion barrier at the fuel/clad interface. This paper gives an overview of the current status ofgives an overview of the current status of U-Mo fuel development, including basic research results, manufacturing aspects, results of the latest irradiations and post irradiation examinations, the approach to fuel performance qualification, and the scale-up and commercialization of fabrication technology. (authors)

58

Material test reactor fuel research at the BR2 reactor  

International Nuclear Information System (INIS)

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 ar2 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 required thermal and hydraulic conditions. The availability of a comprehensive set of post irradiation examination facilities on site complements the versatile BR2 reactor to provide a set of high performance tools for MTR fuel qualification. (author)

59

Advanced research reactor fuel development  

International Nuclear Information System (INIS)

The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ? 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The ?-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content

60

Advanced research reactor fuel development  

Energy Technology Data Exchange (ETDEWEB)

The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

2000-05-01

61

Swelling of U(Mo) dispersion fuel under irradiation – Non-destructive analyses of the SELENIUM plates  

Energy Technology Data Exchange (ETDEWEB)

Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK?CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600 nm Si and 1000 nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ?70% {sup 235}U 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 non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5 × 10{sup 21} f/cm{sup 3}, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article.

Van den Berghe, S., E-mail: sven.van.den.berghe@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Parthoens, Y.; Cornelis, G.; Leenaers, A.; Koonen, E.; Kuzminov, V. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [University of Ghent, Solid State Sciences, Krijgslaan 281, 9000 Gent (Belgium)

2013-11-15

62

Swelling of U(Mo) dispersion fuel under irradiation – Non-destructive analyses of the SELENIUM plates  

International Nuclear Information System (INIS)

Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK?CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600 nm Si and 1000 nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470 W/cm2 peak Beginning Of Life (BOL) power, with a ?70% 235U 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 non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5 × 1021 f/cm3, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article

63

The reprocessing of irradiated fuels improvement and extension of the solvent extraction process; Le traitement des combustibles irradies amelioration et extension du procede utilisant les solvants  

Energy Technology Data Exchange (ETDEWEB)

Improvements made in the conventional tri-butylphosphate process are described, in particular. the concentration and the purification of plutonium by one extraction cycle using tri-butyl-phosphate with reflux; and the use of an apparatus working continuously for precipitating plutonium oxalate, for calcining the oxalate, and for fluorinating the oxide. The modifications proposed for the treatment of irradiated uranium - molybdenum alloys are described, in particular, the dissolution of the fuel, and the concentration of the fission product solutions. The solvent extraction treatment is used also for the plutonium fuels utilized for the fast breeder reactor (Rapsodie) An outline of the process is presented and discussed, as well as the first experimental results and the plans for a pilot plant having a capacity of 1 kg/day. The possible use of tn-lauryl-amine in the plutonium purification cycle is now under consideration for the processing plant at La Hague. The flowsheet for this process and its performance are presented. The possibility of vitrification is considered for the final treatment of the concentrated radioactive wastes from the Marcoule (irradiated uranium) and La Hague (irradiated uranium-molybdenum) Centers. Three possible processes are described and discussed, as well as the results obtained from the operation of the corresponding experimental units using tracers. (authors) [French] On decrit les ameliorations apportees au procede classique utilisant le phosphate tributylique, et notamment la concentration et la purification du plutonium par un cycle d'extraction au tributylphosphate avec reflux, l'utilisation d'un appareillage continu de precipitation d'oxalate de plutonium, de calcination de l'oxalate, et de fluoration de l'oxyde. On presente les modifications envisagees pour le traitement des alliages uranium-molybdene irradies, principalement en ce qui concerne la dissolution du combustible et la concentration des solutions de produits de fission. Le traitement au solvant est egalement utilise pour les combustibles de la pile convertisseuse du plutonium (Rapsodie). On expose et commente le schema du traitement, les premiers resultats experimentaux et le projet d'une installation pilote de 1 kg/jour. L'utilisation de la tn-laurylamine dans le cycle de purification du plutonium est envisagee dans l'usine de traitement de La Hague. On presente le schema adopte et les performances du procede. On envisage la vitrification comme traitement definitif des dechets radioactifs concentres des Centres de Marcoule (uranium-irradie) et La Hague (uranium-molybdene irradie). Trois procedes possibles sont decrits et commentes, ainsi que les resultats d'exploitation des installations correspondantes sur elements traceurs. (auteurs)

Faugeras, P.; Chesne, A. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

64

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts  

International Nuclear Information System (INIS)

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling. In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl3. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K2UCl6 at the uranium surface, within the error of the analytical method

65

Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization  

International Nuclear Information System (INIS)

The replacement of high enriched uranium (U-235 > 85 wt%) by low enriched uranium (U-235 10Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U2Mo compound and alpha-phase, in the heat treated condition. (author)

66

Orientational relationships between phases in the ??? transformations for uranium-molybdenum alloys  

International Nuclear Information System (INIS)

A crystallographic study has been made of the ? ? ? + ? transformation in the alloy containing 3 per cent by weight of molybdenum using electronic micro-diffraction; it has been possible to establish the orientational relationships governing the germination of the ? phase in the ? phase. One finds: (111)? // (100) ?, (112-bar)? // (010) ?, (11-bar 0)? // (001)?. By choosing a monoclinic lattice containing the same number of atoms as the orthorhombic lattice for defining the ? mother phase, the change in structure has been explained by adding a homogeneous (112-bar)? [111]? shearing deformation to a heterogeneous deformation brought about by slipping of the atoms which are not situated at the nodes of this lattice. The identity of the orientation relationships ?/? and ?/?''b and the loss of coherence ? /? as a function of temperature or of time lead to the conclusion that, in the range studied, the ? ? ? transformation begins with a martensitic process and continues by germination and growth. (author)

67

Equipment for decontamination of water contaminated especially with zinc, radium, uranium, molybdenum and oil products  

International Nuclear Information System (INIS)

The design is registered for patent of a decontamination tank which consists of a support tank with a filter unit at the bottom, a sorption cylinder with a conic perforated bottom and a supply, relief and discharge pipe. This single block treatment unit carries out three technological operations: filtration and two sorption steps. The device may be series or parallel connected in columns. The advantage of the device is its mobility and short start-up time. This allows to eliminate accident situations and to treat short-term waste water sources. (E.S.). 1 fig

68

Study of transformations by annealing of the body. Centred cubic ? phase of uranium-molybdenum alloys  

International Nuclear Information System (INIS)

By annealing at different temperatures, we have studied the transformations of the body centred cubic ? phase for two alloys containing 6 and 10 per cent molybdenum by weight respectively. There is a return to the equilibrium state by formation of the stable ? orthorhombic and ? ordered tetragonal phases, following two types of reaction: - pearlite transformation by nucleation and growth from the grain boundaries, preponderant when the annealing takes place at temperature above 400 deg. C, and identical for the two types of alloys. This reaction has already been studied by numerous authors, who have constructed the corresponding TTT curves, - transformation inside the grains of the quenched solid solution when annealing takes place at 400 deg. C or below: 6 per cent alloy - precipitation of fine a phase particles, followed by progressive ordering of the solid solution enriched in molybdenum, 10 per cent alloy - formation of small ordered regions and then a fine a phase precipitate. In the course of this work we have paid particular attention to the study of intragranular reactions after low-temperature annealing, the reactions involved in this case not having been explained up to the present. The ? phase transformation has been studied by means of three techniques: micrography - microhardness tests - X-ray diffraction. (author)

69

Nuclear Safety Considerations in Fabrication of Massive, Partially-Enriched Uranium-Molybdenum Reactor Parts  

International Nuclear Information System (INIS)

Massive metallic components of partially-enriched uranium-235 mixed with 10 wt.% molybdenum have been successfully fabricated at the USAEC Oak Ridge Y-12 Plant for Super Kukla, a prompt burst reactor. Nuclear safety analyses were performed and procedures developed to permit fabrication of the reactor components in the largest single pieces possible within the limitations imposed by criticality and manufacturing capabilities. Metal parts of finished weights up to 268 kg each were cast, machined, inspected and shipped. Nuclear safety problems encountered in the production of approximately 5 tons of these reactor components included considerations of reflected and unreflected massive pieces of uranium metal and alloy, accumulations of machine turnings in various conditions of moderation by hydrogenous liquids and uraniumbearing solutions from plating processes. Although some operational steps were resolved by application of criticality data and established practices for uranium more highly enriched in 235U (? 90%), it was necessary to establish critical parameters for the intermediate 20% enrichment desired and to evaluate the effects of dilution by molybdenum. Calculations to obtain the criticality numbers were made using the Sn reactor transport theory approximation IBM-7090 machine codes DTK and DDK. Hansen-Roach 16 energy group cross-sections were used with appropriate resonance region corrections. Checks against Los Alamos critical experimental data foros Alamos critical experimental data for 28.9, 38.0 and 50.5 % enriched uranium were made to assist in establishing the reliability of the calculations. Each proposed operational step was analysed using the 'double contingency' criterion. On the basis of the analyses, it was possible to devise procedures and equipment to safely allow casting charges of up to 300 kg of uranium metal (60 kg 235U) or 400 kg of alloy (72 kg 235U) in cylindrical crucibles. Especial care was required to prevent inadvertent mixing with either highly enriched uranium or depleted uranium from adjacent working areas. Most of the reactor parts themselves were readily identifiable due to their large size and unique configuration; however, machine turnings, chips and solutions were not sufficiently distinctive for visual identification as 20% enrichment. These materials were accordingly treated as highly enriched (?90%) until proven otherwise by analyses. (author)

70

Fuel flexible fuel injector  

Energy Technology Data Exchange (ETDEWEB)

A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

2015-02-03

71

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009  

International Nuclear Information System (INIS)

This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

72

Electrochemical deposition of chromium and chromium alloys on U-Mo alloy  

International Nuclear Information System (INIS)

Reliable work of fuel elements is achieved by shielding nuclear fuel from the coolant attack, that lessens the fission product release into a nuclear plant coolant loop. Electrochemical deposition of chromium and chromium alloys on the uranium-molybdenum alloy base is one of the shielding methods presented in the paper. Relative simplicity as well as possibility to obtain multilayer coatings on various profile surfaces are obvious asserts of the method. Nickel-chromium and chromium alloy coatings show corrosion resistance during 1000 h at 400 deg C. However electrochemical deposition of uranium-molybdenum alloy doesn't give a reliable shielding

73

Fossil fuels -- future fuels  

Energy Technology Data Exchange (ETDEWEB)

Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

NONE

1998-03-01

74

Fueling systems  

International Nuclear Information System (INIS)

This report deals with concepts of the Tiber II tokamak reactor fueling systems. Contained in this report are the fuel injection requirement data, startup fueling requirements, intermediate range fueling requirements, power range fueling requirements and research and development considerations

75

Fuel assumbly  

International Nuclear Information System (INIS)

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

76

Fuel rods  

International Nuclear Information System (INIS)

Purpose: To prevent fuel can fractures in final burning stage and thereby improve fuel safety by providing a certain gap between a fuel can and fuel pellets in nuclear reactors. Constitution: A plurality of hollow fuel pellets are charged in a fuel can and their upper and lower ends are sealed with end plugs. The gap between the fuel can and the pellets is set within 2.5 - 3.5% of the pellet diameter. As the result, fuel can failures at the final burning stage due to the contact between the fuel pellets and the fuel can are prevented thereby enabling to improve the reactor safety. (Horiuchi, T.)

77

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To prevent failures in fuel elements loaded with fuel pellets containing burnable poisons such as gadolinium. Constitution: A fuel assembly comprises a plurality of fuel elements loaded with fuel pellets consisting only of uranium dioxide and plurality of hollow gadolinium-containing fuel elements loaded with gadolinium-containing fuel pellets prepared by adding gadolinium oxide as burnable poisons to uranium dioxide and formed into hollow cylinder in a fuel can, each disposed in a channel box. Since gadolinium is not contained in the center of the gadolinium-containing fuel pellet, the maximum power for the fuel elements is resulted at an earlier time than that for the solid gadolinium-containing pellets. Since the fuel can rapture distortion is greater as the reactor staying time is shorter, the risk of failures in the gadolinium-containing fuel elements can be reduced as the result. (Horiuchi, T.)

78

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To make uniform the axial load which the fuel rods receive from springs by changing the constraining strength of the springs against the fuel rods in response to the power of the corresponding fuel rods. Constitution: Expansion springs for restraining the axial movements of respective fuel rods are retained between the fuel rods and upper tie plates. The expansion spring is adjusted in its length in response to the power of the fuel rod. Accordingly, the axial load added to the fuel rod can be made uniform, and the deflection of the fuel rod is prevented. (Aizawa, K.)

79

Nuclear fuels  

International Nuclear Information System (INIS)

Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

80

Fuel assembly  

International Nuclear Information System (INIS)

A fuel assembly of a BWR type reactor comprises a rectangular parallelopiped channel box and fuel bundles contained in the channel box. The fuel bundle comprises an upper tie plate, a lower tie plate, a plurality of spacers a plurality of fuel rods and a water rod. In each fuel rod, the amount of fission products is reduced at upper and lower end regions of an effective fuel portion than that in other regions of the effective fuel region. In a portion of the fuel rods, fuel pellets containing burnable poisons are disposed at the upper and lower end regions. In addition, the upper and lower portions are constituted with natural uranium. Each of the upper and lower end regions is not greater than 15% of the effective fuel length. Since this can enhance reactivity control effect without worsening fuel economy, the control amount for excess reactivity upon long-term cycle operation can be increased. (I.N.)

81

Fuel assembly  

International Nuclear Information System (INIS)

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.)

82

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To obtain such fuel rods that all fuel rods in the fuel assembly have the same linear power density without using fuels of several kinds having different concentrations. Constitution: Fuel pellets filled in fuel rods are made in hollow shape. Hollowness is made small at the central part of the assembly, and it becomes larger as it approaches the channel box, whereby the power densities of the corner parts of four fuel rods, where are liable to reach peaks are reduced, and, at the same time, the power density distribution is made uniform as a whole. (Kamimura, M.)

83

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To prevent fuel rod failures due to deposition of scale in spacers, by filling substances with less heat generation to positions corresponding to spacers in a fuel rod. Constitution: A fuel assembly comprises a plurality of fuel rods bundled by spacers, held at the upper and lower positions with tie plates and inserted in a channel. Substances with less heat generation are filled in the fuel rods at the positions corresponding to the spacers. For instance, fuel pellets are filled in lamination within a fuel can and substances with less heat generation such as pellets of uranium at low enrichment degree (natural uranium) are filled and the openings on both ends of the fuel can are sealed with end plugs. This can lower the temperature at the positions corresponding to the spacers in the fuel rod and prevent the deposition of water scales on the spacers. (Horiuchi, T.)

84

Alcohol fuels  

Directory of Open Access Journals (Sweden)

Full Text Available The existing motor fuel alternative, namely alcohol - biomethanol, bioethanol and biobutanol, the possibility of using them in different concentrations of gasoline were consider. From the most perspective of considered alternative fuels for were shows.

?.?. ????????

2010-02-01

85

Fuel gases  

International Nuclear Information System (INIS)

This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

86

Fuel cycles  

International Nuclear Information System (INIS)

AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

87

Fuel element  

International Nuclear Information System (INIS)

A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

88

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To improve the fuel soundness in BWR type reactors by flattening the axial power distribution in fuel assemblies, as well as decreasing the linear power density which restricts the reactor operation. Constitution: Gadolinium oxides (Gd2O3) are added to several of fuel rods in a fuel assembly. The fuel rods incorporated with the gadolinium oxides can satisfy the following conditions: amounts of fission products in the upper parts of fuel rod > amounts of fission products in the lower part of fuel rod, and amounts of gadolinium oxides in the lower part of the fuel rod > amount of gadolinium oxide in the upper part of fuel rod. It also flattens the axial power distribution in the fuel assembly with respect to the axial distribution of uranium 235 and gadolinium oxides in the initial stage of the core and flattens the axial power distribution of the fuel assembly only with respect to the axial distribution of uranium 235 in the finally stage of the core, whereby the axial power distribution of the fuel assembly can be flattened throughout the entire stages of the core. (Moriyama, K.)

89

LPG fuel  

International Nuclear Information System (INIS)

LPG fuel has become frequently used through a distribution network with 2 000 service stations over the French territory. LPG fuel ranks number 3 world-wide given that it can be used on individual vehicles, professional fleets, or public transport. What is the environmental benefit of LPG fuel? What is the technology used for these engines? What is the current regulation? Government commitment and dedication on support to promote LPG fuel? Car makers projects? Actions to favour the use of LPG fuel? This article gathers 5 presentations about this topic given at the gas conference

90

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly comprising an orderly arrangement of fuel rods of a first group containing burnable poisons and fuel rods of a second group not containing burnable poisons, mixed fuels of plutonium oxides and uranium oxides are incorporated to at least a part of the fuel rods in the second fuel rods. The burnable poisons incorporated in the fuel rods of the first group are a mixture of gadolinia (Gd2O3) and europia (Eu2O3). With such a constitution, the reaction rate of MOX fuel rods incorporating the mixed fuels of plutonium oxides and uranium oxides in a thermal neutron region is controlled mainly by Gd, while the reaction rate in an epithermal neutron region is controlled by resonance absorption of Eu in a resonance neutron region. Then, the multiplication factor of the MOX fuel rods can be reduced greatly, thereby enabling to operate and shut down a reactor safely and reliably even if a greater amount of Pu is added to fuel assemblies than usual case. (T.M.)

91

Fuel distribution  

Energy Technology Data Exchange (ETDEWEB)

Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

Tison, R.R.; Baker, N.R.; Blazek, C.F.

1979-07-01

92

Fuel assembly  

International Nuclear Information System (INIS)

The present invention concerns a cluster type fuel assembly for a pressure tube reactor. Most of fuel rods are MOX (mixed oxide) fuel rods. They are divided from the axial top end into three regions, that is, a high Pu enrichment degree fuel region, a low Pu enrichment degree fuel region, a high Pu enrichment degree region. Further, a small number of gadolinium-incorporated enriched uranium fuel rods are present together. They comprise a high concentration gadolinium-incorporated uranium fuel region in the upper portion and a low concentration gadolinium-incorporated uranium fuel region in the lower portion. Accordingly, the power distribution peak in the axial center can be suppressed by the MOX fuel rods when the fuel assembly is loaded. Further, the power distribution peak in the reactor core upper portion can be suppressed in the last stage of the balanced core by fuel rod having two gadolinium concentration regions. This can flatten the reactor core power distribution through out the combustion period. (I.S.)

93

European Fuel Group's fuel performance  

International Nuclear Information System (INIS)

The European Fuel Group (EFG) comprises three member companies and can provide to European facilities products and services which have been developed by the individual members, or jointly. Several advanced fuel features are now being supplied to European plants and this paper offers a summary of the performance of EFG fuel and the background experience of the individual EFG members. (author)

94

Fuel Cells  

DEFF Research Database (Denmark)

Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed.

Smith, Anders; Pedersen, Allan SchrØder

2014-01-01

95

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010  

Energy Technology Data Exchange (ETDEWEB)

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01

96

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010  

International Nuclear Information System (INIS)

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

97

Fuel assembly  

International Nuclear Information System (INIS)

Object: To prevent failure of fuel rods resulting from trouble of coolant loss without modification of basical construction of a cooling device in a low pressure water injection system. Structure: To provide an opening having a sectional area which is 10 to 20% of that of a cooling water passage, at a position lower than an upper tie plate on the wall surface of a channel box enclosing therein fuel rods but at a position higher than an effectively heat-generating upper end of the fuel rods. With this arrangement, flow of spray water into the fuel assembly and resultant impeding phenomenon or the like can be avoided. (Kamimura, M.)

98

Nuclear fuel  

International Nuclear Information System (INIS)

All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.)

99

Fuel rod  

International Nuclear Information System (INIS)

Rod-like material comprising burnable reactivity controlling material, such as Gd2O3 is penetrated through a hollow portion of a plurality of hollow pellets made of UO2 or MOX having holes along the central axes of fuel rods, to provide pellet stacks, and they are inserted to cladding tubes. According to this method, the hollow pellets made of UO2 or MOX can be manufactured in a nuclear fuel production facility in the same manner as in a conventional case. On the other hand, the rod-like material comprising the burnable reactivity controlling materials such as Gd2O3 is previously manufactured in a non-nuclear fuel production facility not in the nuclear fuel production facility, and they can be transported to the nuclear fuel production facility and assembled with the hollow pellets into fuel rods. Accordingly, it is not necessary to dispose two independent systems of a line concerning Gd2O3 and a line concerning only MOX fuel materials in the nuclear fuel production facility, thereby enabling to reduce production cost for nuclear fuel rods. (T.M.)

100

Fuel spacer  

International Nuclear Information System (INIS)

In a fuel spacer, a chamfer is formed on the outside at the lower end of a ferrule. The distance of the flow channel of coolants in a fuel assembly is different for perpendicular direction and diagonal direction between the fuel rods. At a position of the fuel spacer, an annular flow channel is defined with the surface of the fuel rod and the inner side of the ferrule in the perpendicular direction, and an outer side flow channel of the ferrule in adjacent with the annular flow channel is defined in the diagonal direction. In particular, in the annular flow channel in the diagonal direction, the coolant flow having fast flow rate at the periphery of the surface of the fuel rod is introduced to the outer side flow channel by the chamfer then flows in the form of a compressive flow. Since stream lines at the periphery of the surface of the fuel rod do not approach the surface of the fuel rods and the flow is not compressed, the reduction of the thickness of the liquid membrane flow is moderated. This can suppress the occurrence of transient boiling to improve limit power of the fuel assembly. (N.H.)

101

Fuel cells  

Science.gov (United States)

The status of the US Department of Energy's Fuel Cells Program as at the end of FY 85 is described. The report consists of: (1) an overview of the Fuel Cells Program including a brief discussion of how fuel cells work; (2) a synopsis of the Phosphoric Acid Fuel Cell (PAFC), Molten Carbonate Fuel Cell (MCFC), and Solid Oxide Fuel Cell (SOFC) Programs and their 1985 projects; (3) a discussion of the Fuel Cells Advanced Research and Technology Development (AR and TD) Program and projects; and (4) a summary of the Fuel Cells Systems and Applications Program. A common direction of fuel cell development has been to combine individual cells into groups called stacks or modules in order to increase power output. In 1985, the scale-up of PAFC stacks to the 40-kW level continued, and a project involving the manufacturing of 46 power plants was completed. SOFC scale-up proceeded to the 24-cell submodule stage. An MCFC 1-ft stack demonstrated effective management of electrolyte, control of end-cell shorting, and resistance of separator plates to corrosion during 4000 hours of operation. AR and TD provided information on reaction mechanisms and materials for MCFC's, SOFC's, and PAFC's.

102

Fuel taxation  

OpenAIRE

In the autumn of 2000, increases in the price of petrol led to fuel protests across Britain. It was argued that high levels of indirect taxation on fuel, which had risen rapidly in each year from 1993 to 1999 (the "escalator", which saw duties on fuel increase by 3 percentage points above inflation between 1993 and 1997, and 6 points between 1997 and 1999), had provoked the protests. Since abandoning the escalator in the 1999 Pre-Budget Report, the Chancellor has not increased fuel duties abo...

Leicester, A.

2005-01-01

103

Production and Characterization of Atomized U-Mo Powder by the Rotating Electrode Process  

International Nuclear Information System (INIS)

In order to produce feedstock fuel powder for irradiation testing, the Idaho National Laboratory has produced a rotating electrode type atomizer to fabricate uranium-molybdenum alloy fuel. Operating with the appropriate parameters, this laboratory-scale atomizer produces fuel in the desired size range for the RERTR dispersion experiments. Analysis of the powder shows a homogeneous, rapidly solidified microstructure with fine equiaxed grains. This powder has been used to produce irradiation experiments to further test adjusted matrix U-Mo dispersion fuel

104

Fuel assembly  

International Nuclear Information System (INIS)

The present invention provides a suitable flattening for the axial power distribution of a fuel assembly used for a BWR type reactor. That is, at the lower portion of fuels increasing the power at the initial burning stage, fuels containing U-234 instead of U-235 are used without changing the enrichment degree of U-235. Since the void ratio is low and water density is high in the lower portion of the fuels, neutron spectral are soft. On the other hand, due to the difference in the capturing cross section relative to the thermal neutrons, fuels partially containing U-234 instead of U-235 show lower neutron multiplication factor and higher conversion ratio as compared with fuels not containing U-234. Accordingly, by using fuels partially containing U-234 instead of U-235 in their lower portion in which void ratio is lower, the power from the lower protion tending to be increased as compared with that in the upper portion can be reduced. Accordingly, power can be flattened without increasing the enrichment degree in the upper portion of fuels. Further, since the enrichment degree is not increased, reactor shutdown margin is not worsened. (I.S.)

105

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly comprising a water rod in which the height of a liquid level is changed by the increase and the decrease of a coolant flow rate, the number of burnable poison-incorporated fuel rods in adjacent with the water rod in the lower region is decreased than that in the upper region. With such a constitution, the burnable poison-incorporated fuel rods are dispersed and arranged uniformly in the horizontal cross section in the lower region of the fuel assembly. Accordingly, since the infinite multiplication factor in the lower region of the first cycle fuel newly loaded in the reactor in the first stage of the cycle can be suppressed low, the linear power density of the fuel rod can be suppressed within a preferable range. Further, since the combustion rate of the burnable poisons in the lower region of the fuel is made faster than that in the upper region, the lower power peak in the earlier half of the reactor operation cycle is kept longer. Then, nuclear fuel materials can be utilized more effectively. (T.M.)

106

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To reduce useless neutron absorption into a wrapper tube covered a fuel assembly of an LMFBR type reactor by reducing the wall thickness of the wrapper tube. Constitution: The fuel assembly is composed of fuel element clusters and the wrapper tube covering them. The wall thickness of the wrapper tube is changed near the boundary (referred to as transition point) between the core fuel and the blanket fuel. Specifically, the wall thickness of the wrapper tube below the transition point is made uniform determined by the highest inner pressure, which is thinner than the wall thickness above the transition point made uniform determined by the inner pressure at the transition point. By partially decreasing the thickness of the wrapper tube, useless absorption of neutrons into the wrapper tube can be reduced to thereby improve the breeding ratio. (Moriyama, Y.)

107

Fuel assembly  

International Nuclear Information System (INIS)

An obstacle capturing filter interposed between a support lattice and a lower nozzle is constituted by recessing the outer lateral surface of short circular tubes at the outermost edge, among the short circular tubes surrounding a lower end plug of a fuel rod. Since the filter is easily fabricated, it can be manufactured at a reduced cost, and since it forms an obstacle capturing space being isolated from fuel rod cladding tube while allowing thermal expansion and irradiation growth of fuel rods, fretting corrosion of the fuel rod cladding tube caused by obstacles can reliably be prevented. In addition, the wall of the circular tubes as the obstacle capturing filter can be formed thinly, pressure drop of coolant's flow passing therethrough is not increased, and since the outside of the circular tubes at the outer edge is recessed, access thereto can be facilitated in order to detect the amount of the irradiation growth of fuel rods. (N.H.)

108

Fuel cycles  

International Nuclear Information System (INIS)

For the pebble bed reactor there is a large field of variations as to the design of the fuel cycle. This results on one hand from the flexibility in the design of the fuel element and on the other from the possibility of variation in the operation of the reactor. The continuous method of refueling permits the residence time in the reactor of the fuel elements to be optimized without interruption. It further allows mixed charging of the reactor with fuel elements of different design and, if necessary, to recycle discharged elements without reprocessing, which has been demonstrated for 10 years in the Atomic Experimental Reactor AVR. These opportunities are taken advantage of for different fuel cycles to be optimized, compared with each other and for developing the leading facts for the commercial introduction of this reactor. (orig.)

109

Nuclear fuels  

International Nuclear Information System (INIS)

Nuclear fuel refers to any fuel that is consumed or used as the driving force for nuclear energy, most often generated through a fission process where the fuel's atomic elements are forcibly divided in order to produce energy. This fuel typically has to have highly fissionable elements that can absorb neutrons that bombard them in order to be easily split and allow for the harnessing the energy that is produced. Nuclear fuel can also either refer directly to the material that is directly used for the nuclear fission process or the physical objects that are developed from the base fuel and are compositions of both the base material and other elements. The most common base fuels that is used in nuclear reactors are either uranium 235 or plutonium 239, both of which form the backbones of nuclear power generation in modern years Other derivatives of nuclear fuel are used in less common, more contained power generation ways that may not produce as ample amount of energy as the fission process however are generally more contained and safer. This includes the isotope plutonium 238 and other elements that can be used to produce nuclear power through a simple matter of radioactive decay and are very common in atomic batteries and other long-term regular output energy sources. There are a number of other fuel elements that are also used in alternative forms of nuclear power (such as tritium) and can be found as catalysts in the fusion rather than fission process of nuclear poweather than fission process of nuclear power generation where rather than by splitting atoms for energy molecules are forcibly joined together in order to generate power. This process is most common today in hydrogen fuel cells where hydrogen and oxygen are fused to create the byproduct of water while generating electricity. (author)

110

Fuel assembly  

Energy Technology Data Exchange (ETDEWEB)

A fuel assembly of a BWR type reactor comprises a plurality of fuel rods arranged in 9 x 9 matrix and water rods disposed between the fuel rods. The total of the cross sectional area of the water rods is determined as A{sub W}, the cross section of the fuel assembly is determined as A{sub CH}, the total of the cross section of the fuel rods is determined as A{sub rod}, and a wall thickness of the cladding tube of the fuel rod is determined as {delta}. The fuel assembly is constituted so that A{sub rod}/A{sub CH} and A{sub W}/A{sub CH} satisfy the following formulas. (A{sub rod}/A{sub CH}) {<=} -1.050(A{sub W}/A{sub CH}) +0.469, (A{sub rod}/A{sub CH}) {<=} (0.337{delta} + 0.1421)(A{sub W}/A{sub CH}) + 0.0804{delta} + 0.302, 0.98(0.2425{delta} + 0.384)/(0.337{delta} + 1.192) {<=} (A{sub rod}/A{sub CH}) {<=} 1.02(0.2425{delta} + 0.384)/(0.337{delta} + 1.192), 0.98(0.0804{delta} - 0.167)/(-0.337{delta} - 1.192) {<=} (A{sub W}/A{sub CH}) {<=} 1.0(0.0804{delta} - 0.167)/(-0.337{delta} - 1.192). (I.N.)

Masuhara, Yasuhiro; Yokomizo, Osamu; Inoue, Kotaro; Uchikawa, Sadao; Aoyama, Motoo; Yamashita, Jun-ichi; Yoshimoto, Yuichiro; Yasuda, Tetsuo; Hirakawa, Hiromasa

1998-07-31

111

Fuel assembly  

International Nuclear Information System (INIS)

A rotatable cylinder having an opening that gradually deviates in one direction toward upward relative to the central axis of a fuel assembly is disposed at the inside of a handling head of a fuel assembly. The cylinder is disposed so that the upper portion of the opening faces toward adjacent control rod assembly. In the fuel assembly, coolants flow out along the deviated opening of the handling head. With such a constitution, a mixed region of high temperature coolants flown out from the fuel assembly and low temperature coolants flown out from the adjacent control rod assembly is enlarged downwardly. Accordingly, temperature fluctuation of coolants in the upper portion of the reactor core can be suppressed. As a result, thermal impacts caused not only a connection portion between a guide tube of control rod drives and a lattice plate but also optional portions of the upper structures of a reactor core are moderated. (I.N.)

112

Fuel cell  

International Nuclear Information System (INIS)

A fuel cell construction of economical design is disclosed. In the construction, a honeycomb separator is used to define a plurality of compartments which are separated from one another by a porous cell wall. Electrolyte is provided in the cell walls while alternate compartments of the cell contain either an oxidant or a fuel for the fuel cell. The cells contain suitable electrochemical catalyst materials on the walls thereof and electrode structures in the cells so that the oxidation of the fuel may take place in the electrolyte found in the cell walls in order to generate current for the cell. In accordance with preferred teachings, the separator is an extruded ceramic material such as used for the substrate of automotive catalytic converters

113

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly for high burnup degree in which large-diameter water rods are disposed at the central portion for flattening the power distribution at a cross section, the distribution of burnable poisons is made optimal while considering an axial void coefficient distribution. That is, since the power in the vicinity of the large-diameter water rod is lowered at a high void coefficient in the upper portion of the reactor core, compared with that in other fuel regions, the concentration of the burnable poisons is made lower than that in other fuel regions. Since the homogenity of the assembly is improved at a low void coefficient in the lower portion of the reactor core, the difference of the concentration of the burnable poisons is reduced. Thus, the axial distribution of the fuel assembly is flattened after the middle stage of the operation cycle and a sufficient scram reactivity can be ensured, to improve the reactor safety. (N.H.)

114

Fuel cells  

Directory of Open Access Journals (Sweden)

Full Text Available The current state of development of fuel cells as potential power sources is reviewed. Applications in special fields with particular reference to military requirements are pointed out.

D. N. Srivastava

2014-05-01

115

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To enable a further effective utilization of uranium resources in the fuel assembly for BWR type reactors. Constitution: Fuel rods in a fuel assembly are grouped into four small assemblies either of 5 x 3 matrix or 3 x 5 matrix and contained within a channel box. A non-boiling water flow channel is disposed at a central region of the fuel assembly and light water is caused to flow therethrough. Each of the four small assemblies is adapted such that it can be turned by 180 0 by a predetermined spacer and inserted into the original position. Since the non-boiling water region is present, thermal neutron fluxes can be increased to improve the burnup efficiency. Further, since the uranium remained in the central region can be burnt at the peripheral region by turning the direction of the small assembly, the effective utilization of uranium resource is possible. (Horiuchi, T.)

116

Fuel behaviour  

International Nuclear Information System (INIS)

A short summary of the main critical points in fuel performance of nuclear power reactors from chemical and mechanical point of view is given. A schedule for a limited research program is included. (author) 17 refs

117

Fuel assembly  

International Nuclear Information System (INIS)

In a horizontal cross section in the vicinity of the upper end portion of short fuel rods, the enrichment degree of fuel rods at the outermost circumferential corner is decreased to less than 60% relative to the average enrichment degree in the horizontal cross section, to reduce a cold temperature power peaking in the outermost circumferential corners. Next, in the second layer from the outer circumference in a 9 x 9 arrangement, gadolinia-incorporated fuel rods are disposed each at a position in adjacent with the corner, to reduce the cold temperature power peaking for the fuel rods other than the corners. Then, the cold temperature power peaking is made substantially the same value as that during power operation, so that local increase of neutron fluxes is suppressed to suppress increase of the maximum fuel rod enthalpy upon control rod dropping accidents at reactor startup. As a result, a margin relative to a threshold value causing a fuel rod failure can be increased. (N.H.)

118

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To enable to flatten the power such that the local power peaking can be reduced even upon using uranium of only one degree of enrichment. Constitution: A plurality of fuel rods filled with fissile material to the inside thereof and a plurality of fuel rod filled with the same fissile material and burnable poisons to the inside thereof are arranged such that the transversal cross section of a bundle constituted with their regular arrangement forms a square configuration as a whole. Hollow pellets each having a hollow portion with a diameter about 1/2 - 1/3 of the outer diameter are used as the pellets to be charged within those fuel rods situated at the corners of the square configuration. Accordingly, since fuel rods charged with hollow pellets are employed to a portion of the fuel assembly, the local power peaking can be reduced as compared with the prior case in an advanced burning stage to thereby improve the fuel safety. (Yoshihara, H.)

119

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To increase the charging fuel quantity and at the same time increase the degree of burn-up by inserting a control rod, which has a cladding tube filled with a moderator and burnable poison, into the fuel assembly. Constitution: A control rod, which has a cladding tube filled with burnable poison such as gadolinium, gadolinium compounds, boron, boron compounds and so forth and a moderator such as light water, heavy water, beryllium, beryllium compounds, carbon and carbon compounds, is inserted in lieu of a suitable number of fuel pins into the fuel assembly. Since the burnable poison is charged into the cladding tube without mixing it with the moderator, it is possible to increase the charging quantity of the burnable poison compared to the former system where fuel is mixed with the burnable poison. Thus, the effect of increasing the reactivity is increased, so that it is possible to increase the charging quantity of fuel and the degree of burn-up to that extent. (Horiuchi, T.)

120

Fuel assembly  

International Nuclear Information System (INIS)

In a nuclear assembly having fuel rods arranged in a triangle lattice-like configuration, if its horizontal cross section is divided into a plurality of triangle regions by linear segments connecting the center with each of the corners of the fuel assembly, water rods having unsaturated water filled therein are disposed at the position for the gravitational center of each of the triangle regions or the position for the center of the fuel assembly. Since the ratio of fuel rods in adjacent with the unsaturated water is increased, the distribution of the number ratio of hydrogen atoms to uranium atoms is made uniform, so that the thermal neutron flux distribution and the power distribution are flattened. Further, since the change of the number ratio of hydrogen atoms to uranium atoms at the periphery of fuel rods is decreased when the void ratio is changed, the absolute value of the void reactivity coefficient is decreased. Throughout the operation period, the power distribution in the fuel assembly can be flattened, thereby enabling to decrease the maximum uranium enrichment degree. Further, the absolute value of the void reactivity coefficient can be decreased and the development of transient events can be moderated. (N.H.)

121

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To ensure a required reactor shutdown margin even if the enrichment degree is increased in fuel assemblies use in BWR type reactors so that the reactor available factor and the fuel economy may not be reduced. Constitution: Since voids are generated in the water within the reactor core to generally worsen the neutron deceleration in the upper portion and reduce the reaction ratio between nuclear fuel materials and neutrons in a BWR type reactor. Therefore, the amount of fuel assemblies not yet burnt is increased in the upper portion of the reactor core. Then, in order to suppress the excessive increase in the reactivity in the upper portion of the reactor core, a high concentration region of the burnable poisons is formed in the upper portion of the reactor core to ensure the reactor shutdown margin efficiently even for the fuel materials with high enrichment degree. Further, since the excess burnable poisons are distributed not over the entire upper portion of the reactor core, the presence of the burnable poisons at the final stage of the burning cycle can be suppressed to decrease the reactivity loss and reduction in the reactor available factor and the fuel economy due to the shortening of the cycle length. (Kawakami, Y.)

122

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To obtain both high burnup degree and suitable reactor-shutdown margin in view of the safety. Constitution: The gadolinia concentration in the low gadolinia concentration region near the upper end of the fuel filling region is lowered by 1.5 % by weight as compared with that in the high gadolinia concentration region therebelow in the first fuel rod. Then, the gadolinia concentration in the low gadolinia concentration region near the upper end of the fuel filling region in the second fuel rod is also lowered by about 1.5 % by weight as compared with that in the high gadolinia concentration in the first fuel rod. This shortens the time for burning out gadolinia and the infinite multiplication factor starts to reduce at the initial stage of the operation cycle. As a result, the reactivity suppressing effect due to the residue of gadolinia is reduced to increase the operation period by about 20 days as compared with the usual case and, as a result, fuel economy can be improved. (Yoshihara, H.)

123

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly to be charged in a BWR-type reactor, the dimension of the outer width of a channel box is changed without changing the dimension of the inner width thereof in vertical direction. The wall thickness of the channel box is increased at the upper or the lower portions of the fuel assembly, whereas, it is reduced at the lower or the upper portion of the fuel assembly. During a predetermined period from the initial stage of the fueling, a portion of the channel box with the increased thickness is set at the upper portion. With such a constitution, since an aqueous region is reduced to increase the void efficiency in the upper portion of the fuel assembly, fission reaction can be suppressed in the upper portion thereof. Accordingly, 239Pu in the upper portion of the fuel assembly is accumulated with scarce nuclear fission. The portion of the channel box with the increased thickness is set in the lower portion, to effectively burn the 239Pu for a remaining period. With such procedures, spectral shift operation is enabled, to improve reactivity and burnup degree. (T.M.)

124

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To improve the burning efficiency by improving the infinite multiplication factor of a fuel assembly. Concentration: Water rods sealed with a great amount of moderators instead of nuclear fission materials are disposed at the central portion, the inside of the assembly is divided into three regions, i.e., the first region at the first circumferential position, the second region at the second circumferential position and the third region inner from the first and second regions from the outermost circumference of the fuel rod, and the average enrichment degree of nuclear fuel materials of the fuel rod in each of the regions is made such that the enrichment degree is higher in the first region than in the second region, and that in the third region is equal to or higher than that in the first region. Accordingly, since the infinite multiplication factor of the assembly can be improved, burn-up degree upon recovery of the fuel is increased, enabling to utilize more energy and the burning efficiency for fuel can be improved. (Yoshihara, H.)

125

Candu fuel and fuel cycles  

International Nuclear Information System (INIS)

A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous emissions are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. The technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy. The world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, CANDU reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and ealmost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuels which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the CANDU reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential CANDU fuel cycle developments can be accommodated in existing reactor designs, allowing operation today on currently available fuels and switching to other fueling options as market conditions change. This establishes an important freedom from future resource constraints without depending on future commercialization of challenging and expensive technologies such as fast breeder reactors, yet, once these are commercially available, CANDU and fast breeder fuel cycles are complementary and can achieve a highly advantageous synergism. This paper examines the fuel cycle option which CANDU reactor technology can accommodate, including the use of slightly enriched uranium direct use of spent pressurized water reactor fuel in CANDU (dupic), burning recovered uranium, mixed plutonium and uranium oxides or actinides and the use of thorium based fuel cycles. These options provide CANDU reactors with the most flexible fuelling of any reactor type, which are readily adaptable to meeting future variations in energy markets, regardless of what these may be. (author)

126

Nuclear fuels  

International Nuclear Information System (INIS)

This seminar will help in understanding us about the nuclear fuels. Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Most nuclear fuels contain heavy fissile elements that can be made to undergo a nuclear fission chain reaction in a nuclear reactor. The most common fissile nuclear fuels are 235U and 239Pu. The actions of mining, refining, purifying, using, and ultimately disposing of these elements together make up the nuclear fuel cycle. Not all nuclear fuels are used in fission reactors. Plutonium-238 and some other elements are used to produce small amounts of nuclear power by radioactive decay in radioisotope thermoelectric generators and other atomic batteries. Light nuclides such as 3H (tritium) are used as fuel for nuclear fusion. uranium-238 is fissionable but will not sustain a neutron chain reaction. Neutrons produced by fission of e.g. 235U have an energy of around 2 MeV and only a minority have enough energy to cause fission of 238U, but neutrons produced by the deuterium-tritium fusion reaction have an energy of 14.1 MeV and can effectively fission 238U and other non-fissile actinides. Enriched uranium is a kind of uranium in which the percent composition of uranium-235 has been increased through the process of isotope separation. Natural uranium is 99.284% 238U isotope, with 235U only constituting about 0.711% of its weight. 235U is the only isotope existing in nature (in any appreciable amount) that is fissile with thermal neutrons. There are about 2,000 tonnes (t, Mg) of highly enriched uranium in the world, produced mostly for nuclear weapons, naval propulsion, and smaller quantities for research reactors. The 238U remaining after enrichment is known as depleted uranium (DU), and is considerably less radioactive. At present, 95% of the world's stocks of depleted uranium remain in secure storage. Fissile nuclides in nuclear fuels include: a) Uranium-235 which occurs in natural uranium and enriched uranium, b) Plutonium-239 bred from uranium-238 by neutron capture, c) Plutonium-241 bred from plutonium-240 by neutron capture, d) Uranium-233 bred from thorium-232 by neutron capture. (author)

127

Fuel rods  

International Nuclear Information System (INIS)

Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

128

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To increase the average reactor core reactivity and improve the fuel economy by properly arranging the content of nuclear fissile material in a fuel assembly. Constitution: The content of nuclear fissile material (uranium 235) in the middle region is made greater than that in the upper and lower regions of a fuel assembly. For instance, the reactor core height is equally divided into 24 parts in the axial direction of the fuel assembly, while expressing the lower end part as 0 and the upper end part as 24. Then, the enrichment degree of uranium 235 is set to 1.33 wt% at the parts 0 - 2 and 23 - 24, 2.08 wt% at the parts 2 - 8 and 21 - 23, and 2.25 wt% at the parts 8 - 21 of the axial height. In such a distribution, since the enrichment degree of uranium 235 is reduced in the upper and lower regions of the reactor core, the content of burnable poisons, for example, gadolinia in these regions can be reduced as compared with that in the middle region of the reactor core, thereby the residual amount of the burnable poisons can be decreased at the final stage of the fuel cycle. (Horiuchi, T.)

129

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly, first spacers having relatively low pressure loss are disposed at the uppermost stage and second spacers which causes relatively less boiling transient are disposed for total four stages including the second stage from the uppermost stage. Since a pressure loss multiplication coefficient of two phase flow is great in the upper portion of the fuel assembly and contribution of the spacers to the pressure loss for the entire fuel assembly is great, spacers having low pressure loss are disposed. Further, when the thickness of the water membrane on the surface of fuel rod is 0, heat removal due to evaporation of water is reduced and a temperature on the surface of the fuel rod is abruptly elevated to cause boiling transient. Spacers having relatively great water membrane forming effect to cause less boiling transient are disposed at a position where the thickness of the water membrane undergoes an effect of spacers greatly, in other words, at a position higher than about 13/24 and lower than about 22/24 from the lower portion where coolant flow becomes circular two phase flow. Critical power characteristics can thus be maintained and improved. (N.H.)

130

Fuel assembly  

International Nuclear Information System (INIS)

In a BWR type reactor, the power distribution in the axial direction of a reactor core is low at an upper portion and high at a lower portion, and burnable poisons are used not only for the control of excess reactivity but also for the control of power distribution in axial direction of the reactor core. That is, it is necessary to lower the axial power peaking at the initial stage of operation cycle in order to further improve the thermal characteristics such as linear power density and nuclear thermohydrodynamic stability. In view of the above, in a fuel assembly, fuel rods having the lowest burnable poison concentration at the lower region are spaced apart from other burnable poison-incorporated fuel rods, and fuel rods not containing burnable poisons are disposed between them. Thus, the degree of downwardly convexed neutron multiplication factor upon their rise caused by the effect of the burnable poisons in the lower region of the burnable poison-incorporated fuel rods can be sufficiently compensated by the poisons at the lowest concentration. The axial power peaking at the initial stage of the operation cycle can be effectively suppressed. (N.H.)

131

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To maintain the axial power distribution always flat in a BWR type reactor by making infinite multiplication factors in the upper portion greater than that in the lower portion of fuel assemblies. Constitution: Fuel assemblies are substantially divided axially into two regions i.e., an upper region and a lower region, in which the degree of enrichment in the upper region is made higher than that in the lower region so that the decrease in the infinite multiplication factors due to the void in the upper portion of the reactor core may be compensated during reactor operation where the fuel assemblies are mounted to the reactor core. This enables to maintain the axial power distribution always flat in a simple control rod operation by using, as control rods, only deep control rods without using shallow control rods and specific gadolinia distribution. (Seki, T.)

132

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To effectively utilize fuel and flatten the axial power distribution by reducing the difference in the infinite multiplication factors between the central part and the peripheral part in a plane perpendicular to the shaft of the fuel assembly in the upper region to a value less than that in the lower region. Constitution: The difference in the infinite multiplication factors between the central part and the peripheral part in a plane perpendicular to the shaft of the fuel assembly in the upper region is made smaller than the difference in the infinite multiplication factors between the central part and the peripheral part. Such an organization can be realized by, for example, changing the concentration degree distribution or changing the quantity of gadolinium added. Better effects can be attained, when such a structure that the mean concentration in the upper region is made larger than that in the lower region is concurrently used with the above described structure. (Yoshihara, H.)

133

CANDU fuel  

International Nuclear Information System (INIS)

The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

134

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly, a nuclear fuel pellet of a short fuel rod is formed to a hollow shape. Then, the temperature elevation of the pellet is decreased and the emission of FP gases is reduced. The effect of lowering the temperature of the pellet depends on the size of the diameter of a central hole and, as the diameter of the hole is increased, the temperature of the pellet is lowered. However, if the diameter of the hole is too great, since the amount of fissionable products charged in the control rod is rather decreased, it is not economical in view of the reactor core performance. Specifically, it is desirable that the size of the central hole is 3 to 15% of the volume of the pellet. The length of the plenum portion is shortened without excessively increasing the inner pressure of the fuel rod, and the amount of the thermal neutrons absorbed by the plenum portion is saved without deteriorating the safety, to improve the neutron economy. (N.H.)

135

Fuel assembly  

International Nuclear Information System (INIS)

Fuel rods are arranged by required number in a square lattice together with water rods in the central portion, upper and lower ends thereof are mounted to upper and lower tie plates, middle portions are supported by a plurality stage of spacers and a channel box is put over the outer circumferential surface thereof. All of the fuel rods are constituted with natural uranium or uranium at low concentration similar thereto, and burnable poisons such as gadolinia (Gd2O3) and boron carbide (B4C) are incorporated to a portion of fuel rods entirely or partially in the axial direction. Since this can substitute a portion of excess reactivity with the fuel assembly, it is possible to reduce the amount of the control rods used and the radiation amount to the control rods. Further, since the amount, the kind and the positions to be placed of poisons can properly be selected, flexibility can be provided to the reactor core design. (T.M.)

136

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To increase the reactivity at the final burning stage and increase the burning degree by controlling the power distribution in the direction of the reactor core height in BWR type reactors. Constitution: Fuel assembly is divided into two regions along the direction of the reactor core height such that the difference between the average uranium enrichment degree of the central fuel rods and the average uranium enrichment degree in the peripheral fuel rods in the upper region is made larger than the difference in the lower region. As a result, higher reactivity in the lower region and the lower reactivity in the upper region can easily be attained at the initial burning stage, whereby the power distribution curve is expanded downwardly, the burning at the lower portion of the reactor core is promoted, gas bubble generation portions are situated at the lower portion of the reactor core and the plutonium accumulation is increased. As the operation advances, since fuel burning takes place rapidly in the lower portion of the reactor core, the power ratio is decreased, the gas bubble generation portions are situated higher, the void coefficient in the reactor core is reduced, the reactivity is increased and the burning degree is increased together with plutonium accumulation at the final burning stage. (Moriyama, K.)

137

Fuel assembly  

International Nuclear Information System (INIS)

Heretofore, as a means for attaining flattening of the axial power distribution, a method of axially changing the fuel enrichment degree or a method of axially changing the amount of burnable poisons has been adopted. However, the former has a problem in that the difference of a neutron multiplication factor between the operation state and the cold state, and the latter has a problem of causing incomplete burning. Then, in the present invention, fuels containing Np-237 is partially used instead of U-238 without changing the enrichment degree of U-235 in the lower portion of the fuels where the power is increased at the initial stage of burning. With such a constitution, the axial power distribution can be flattened without increasing the enrichment degree in the upper portion. Further, since the enrichment degree in the upper portion is not increased, the reactor shutdown margin is not worsened. Further, since the neutron multiplication factor is lowered at the initial stage of the burning by using Np-237 in the lower portion of the fuels, it is useful also to the reduction of excess reactivity. (T.M.)

138

Abstracts and papers of the 1999 International RERTR Meeting  

International Nuclear Information System (INIS)

The papers presented at the 22nd International RERTR Meeting dealt with the following topics: development and testing of new fuel elements (uranium-molybdenum alloys); research reactors core conversion studies (change from highly to moderately or slightly enriched uranium), including both measurements and calculations: spent fuel storage and transportation; production of 99Mo from low enriched uranium. A number of papers were devoted to the status and future of national RERTR programs

139

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To decrease the required amount of uranium in BWR type reactors by improving the conversion ratio and effective burning of nuclear fission materials. Constitution: The reactor core and the fuel assembly are equally divided into upper and lower two stages in the present invention. The fuel assembly is placed in the upper stage of the reactor core for the first one-half period and then placed at the lower reactor core in the remaining one-half period. Since the burn-up degree is smaller at the upper stage of the reactor core and greater at the lower stage of the reactor core, the reactivity is greater in the upper stage than in the lower stage, thereby causing the tendency that the power distribution is distorted greatly upwardly in the reactor core. However, the reactivity is suppressed by the high void ratio at the upper stage portion to thereby attain a burner reactor core having high conversion ratio. (Kamimura, M.)

140

Fuel recycling  

International Nuclear Information System (INIS)

The global nuclear power programme should be designed not only to produce electricity at the lowest possible cost, but also to make the best use of our fissile fuel reserves in the longer term. This clearly indicates the necessity of using breeder reactors, which with plutonium recycling, can achieve total fuel utilisation figures of 70% to 80% as opposed to the very small percentages available from non-breeders, even with recycling. The plutonium can be separated from spent fuel elements chemically. The United Kingdom is in a favourable situation to initiate a fast breeder reactor programme because it has appreciable supplies of plutonium accumulated from the Magnox programme, but on a global scale there is a danger that a sudden expansion of the nuclear programme based on non-breeder reactors will exhaust the supplies of commercially viable uranium before adequate supplies of plutonium have been built up to provide the cores for a significant fast breeder programme. This situation will be worse if, as seems likely, the thermal programmes are based on reactors which are poor producers of plutonium and themselves require enriched fuel. A more modest global expansion of the thermal nuclear programme to about 600,000 MW(e) by the year 2000 is possible using reasonably economic uranium reserves. If this were based on thermal reactors which were reasonably good producers of plutonium (Candu, Magnox and HTR) the programme could provide plutonium for the cores of 800, 000 Mvide plutonium for the cores of 800, 000 MW(e) of installed fast breeder capacity by the year 2000. Thereafter, if the doubling time of the electricity demand is shorter than that of the plutonium inventory, either the gas-cooled fast breeder or a combination of thermal 'near breeders' and liquid-metal fast breeders will be required. (author)

141

Fuel assembly  

International Nuclear Information System (INIS)

An inlet nozzle having a plurality of inlets for coolants is disposed at the lower end of a fuel assembly, and a handling head is disposed at the upper end. The inside of the opening of the handling head is enlarged upwardly. Coolants flown out from assemblies having small heat generation value, such as a control rod assembly and a blanket fuel assembly, have a lower temperature compared with coolants flown out from a fuel assembly. High temperature coolants flown out from the handling head and low temperature coolants flown out in adjacent therewith uprise toward upper structural components under mixing above the reactor core. Since the inside of the opening of the handling head is enlarged upward the mixing region of the high temperature coolants and the low temperature coolants is enlarged downward. Accordingly, temperature fluctuation of coolants above the reactor core is suppressed. With such a constitution, thermal shocks are moderated and thermal fatigue exerted on the upper structural components is moderated. (I.N.)

142

Fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To provide fuel assemblies suitable to the increase of the reactor thermal margin and to the improvement in the nuclear thermo-hydrodynamic stability with no remarkable change for the design of fuel assemblies. Constitution: Spacers, constituted such that the pressure loss is made larger at the central portion and smaller at the periphery of the coolant flow channel cross section, are disposed in a region between 3/4 - 4/4 of the whole length from the lower end of the fuel rod, in which the 2-phase stream can take an annular flowing mode. In this way, the tendency of decreasing the coolants at the peripheral area due to the presence of the channel box is offset by the effect resulted by the disposition of the spacers and the effect of disturbing the gas phase flow, by which the defoliation of the coolant liquid film on the cladding tube is suppressed and sufficient heat removal is conducted with the liquid film to increase the thermal margin. (Kamimura, M.)

143

Fuel assembly  

International Nuclear Information System (INIS)

In a spectral shift operation, the difference of water density along with the flow rate change is decreased, and no sufficient improvement can be obtained for fuel economy. That is, a lower limit for the reactor core flow rate is restricted in view of stability, while an upper limit is imposed by the stroke of a reactor core recycling pump. Then, the flow rate in a coolant inlet hole of a water rod is controlled in accordance with the flow rate of coolants flown in a channel box. That is, the water density is kept low by realizing a higher void state during low flow rate period till the last stage of a cycle, and to form and accumulate plutonium 239. The water density is increased under a low void state during high flow rate period in the last stage of the cycle where the reactivity is lowered, to burn plutonium 239. Accordingly, since the spectral shift operation performance is improved to further improve fuel economy and also improve the limit power thermal margin of fuels can be improved. (N.H.)

144

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly for a BWR type reactor, a moderator guide tube is disposed at the central axial direction and a hollow guide tube having a cross like-cross section is disposed at the outerside of the moderator guide tube. Moderators flowing down in the moderator guide tube are discharged from a discharge port to the lower periphery of fuel rods, thereby making the flow channel resistance maximum when the reactor core flow rate is minimum. On the other hand, a flow channel resistance body is disposed at the inlet of the moderator guide tube to reduce the flow channel resistance continuously along with increase of the flow rate. At the initial stage of the operation cycle, the flow rate of the moderators introduced from the discharge port to the axial lower portion is decreased and the void ratio at the upper portion is increased to harden spectrum and form and accumulate a great amount of plutonium. On the other hand, at the final stage of the operation, the void ratio at the lower portion is decreased and the plutonium accumulated at the initial stage of operation cycle is burnt. Thus, the flow channel resistance in the moderator guide tube can be continuously controlled by the flow channel resistance body in accordance with the reactor core flow rate to improve the fuel economy. (N.H.)

145

Maps showing distribution of pH, copper, zinc, fluoride, uranium, molybdenum, arsenic, and sulfate in water, Richfield 1 degree by 2 degrees Quadrangle, Utah  

Science.gov (United States)

These maps show the regional distribution of copper, zinc, arsenic, molybdenum, uranium, fluoride, sulfate, and pH in surface and ground water from the Richfield 1° x 2° quadrangle. This study supplements (Miller and others, 1984a-j) the regional drainage geochemical study done for the Richfield quadrangle under the U.S. Geological Survey’s Conterminuous United States Mineral Assessment Program (CUSMAP). Regional sampling was designed to define broad geochemical patterns and trends which can be used, along with geologic and geophysical data, to assess the mineral resource potential of the Richfield quadrangle. Analytical data used in compiling this report were published previously (McHugh and others, 1981). The Richfield quadrangle in west-central Utah covers the eastern part of the Pioche-Marysvale igneous and mineral belt that extends from the vicinity of Pioche in southeastern Nevada, east-northeastward for 250 km into central Utah. The western two-thirds of the Richfield quadrangle is in the Basin and Range Province, and the eastern third in the High Plateaus of Utah subprovince of the Colorado Plateau. Bedrock in the northern part of the Richfield quadrangle consists predominantly of latest Precambrian and Paleozoic sedimentary strata that were thrust eastward during the Sevier orogeny in Cretaceous time onto an autochthon of Mesozoic sedimentary rocks in the eastern part of the quadrangle. The southern part of the quadrangle is largely underlain by Oligocene and younger volcanic rocks and related intrusions. Extensional tectonism in late Cenozoic time broke the bedrock terrane into a series of north-trending fault blocks; the uplifted mountain areas were deeply eroded and the resulting debris deposited in the adjacent basins. Most of the mineral deposits in the Pioche-Marysvale mineral belt were formed during igneous activity in the middle and late Cenozoic time.

McHugh, J.B.; Miller, W.R.; Ficklin, W.H.

1984-01-01

146

Transport fuel  

DEFF Research Database (Denmark)

Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds. Advanced biofuels based on forest biomass are not yet being produced on a large scale, but are expected to have a better life-cycle emission profile than conventional biofuels. The pathways from feedstock to advanced biofuel are diverse in respect to capacity, technology and final product. Three promising conversion technologies are presented below: pyrolysis, biochemical conversion and gasification

Ronsse, Frederik; JØrgensen, Henning

2014-01-01

147

Fuel assembly  

International Nuclear Information System (INIS)

Object: To reduce the quantity of deformation due to enlargement of a channel box and a prevent leakage of the coolant. Structure: A plate spring for preventing the leakage of a coolant and to be inserted in the space defined between a lower tie plate for supporting the fuel rod and channel box covering the periphery of the tie plate, is folded in a number of stages in the vertical direction and is provided with alternate protuberances to be in forced contact with said tie plate and channel box. (Aizawa, K.)

148

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To maintain the reactivity of the reactor core to a level above a required value for a long period of time to thereby enable long time operation of the reactor. Constitution: The distribution of the total amount of burnable poisons in the upper-half and the lower-half of the fuel rods incorporated with burnable poisons are made such that the reactor-core power distribution forms the lower-peak type at the initial and the medium stages in the burning operation period, and the upper-peak type at the final stage of the period. For instance, two out of eight burnable poison fuel rods are charged with pellets of uranium dioxide incorporated with about 5%(by weight) of gadolinia in the upper portion and pellets of uranium dioxide incorporated with about 1%(by weight) of gadolinia in the lower portion. The remaining six rods are charged with pellets of uranium dioxide incorporated with about 5%(by weight) of gadolinia along the whole length thereof. (Sekiya, K.)

149

Fuel assembly  

International Nuclear Information System (INIS)

A method of increasing the changes of the reactor core flow rate or the axial localization of the power distribution is used for greatly changing void coefficient in a reactor core at the early and final stages of an operation cycle in order to utilize nuclear fuels effectively. However, it involves a limit in the stability and the integrity. In view of the above, a water rod loaded has such a structure that coolants enter from a lower inlet port and flow out from an upper exit port, in which the flow channel is finely divided in the upper portion and the number of the channels in made greater than that in the lower portion and, further, the hydraulically equivalent diameter in each of the channels is decreased. Water level is formed at a low position in the water rod in the early stage, while water of liquid phase is flown out from the upper exit port and fill the inside of the water rod with water of liquid phase in the final stage. As a result, since the void coefficient in the reactor core can be changed greatly between the early and the final stages, the nuclear fuel materials can be utilized effectively. (N.H.)

150

Fuel assembly  

International Nuclear Information System (INIS)

High temperature adsorbents for adsorbing radioactive nuclides in the reactor atmosphere (for example, Fe-Ti-O ceramics) are used as main materials for upper and lower nozzles and upper and lower tie plates as the constituent elements of a fuel assembly. Alternatively, the adsorbents are coated to the constituent elements. Thus, metal ions (Co, Ni, Mn, etc.) are held on the surface of the high temperature adsorbents by means of chemical adsorbing reactions under chemical bonding force. Further, those adsorbed in high temperature water are not easily leached again since extremely strong irreversible chemical adsorbing reaction is taken place. Accordingly, the amount of radioactive nuclides contained in recycled water can be decreased without additionally disposing the facility and reducing the thermal efficiency. Further, it is possible to decrease the exposure dose of operators working at the periphery of recycling water pipeways, etc. (T.M.)

151

RERTR-13 Irradiation Summary Report  

Energy Technology Data Exchange (ETDEWEB)

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

D. M. Perez; M. A. Lillo; G. S. Chang; D. M. Wachs; G. A. Roth; N. E. Woolstenhulme

2012-09-01

152

Fuel Cell Animation  

Science.gov (United States)

This fuel cell animation demonstrates how a fuel cell uses hydrogen to produce electricity, with only water and heat as byproducts. The animation consists of four parts - an introduction, fuel cell components, chemical process, and fuel cell stack.

US Department of Energy - Office of Energy Efficiency and Renewable Energy - Energy Education and Workforce Development

153

Thorium fuel cycle management  

International Nuclear Information System (INIS)

In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

154

Fuel Cell Demonstration with Liquid Fuels  

Energy Technology Data Exchange (ETDEWEB)

This article explains some of the advantages of using the somewhat old-fashioned 'dissolved fuel' fuel cell in an educational context. It permits the use of a wide range of convenient liquid fuels and allows the fuel cell design to be very simple. This makes for a fuel cell that is low in cost, cheap to run, and extremely simple to use. (author)

Larminie, James [Electro-Chem-Technic, 81 Old Road, Headington, Oxford, OX3 7LA (United Kingdom)

2000-07-01

155

Constant strength fuel-fuel cell  

International Nuclear Information System (INIS)

A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

156

Fuel-cycle costs for alternative fuels  

International Nuclear Information System (INIS)

This paper compares the fuel cycle cost and fresh fuel requirements for a range of nuclear reactor systems including the present day LWR without fuel recycle, an LWR modified to obtain a higher fuel burnup, an LWR using recycle uranium and plutonium fuel, an LWR using a proliferation resistant 233U-Th cycle, a heavy water reactor, a couple of HTGRs, a GCFR, and several LMFBRs. These reactor systems were selected from a set of 26 developed for the NASAP study and represent a wide range of fuel cycle requirements

157

Fuel element loading system  

International Nuclear Information System (INIS)

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

158

Fuel processors for fuel cell APU applications  

Science.gov (United States)

The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

159

Fuel processor for fuel cell power system  

Science.gov (United States)

A catalytic organic fuel processing apparatus, which can be used in a fuel cell power system, contains within a housing a catalyst chamber, a variable speed fan, and a combustion chamber. Vaporized organic fuel is circulated by the fan past the combustion chamber with which it is in indirect heat exchange relationship. The heated vaporized organic fuel enters a catalyst bed where it is converted into a desired product such as hydrogen needed to power the fuel cell. During periods of high demand, air is injected upstream of the combustion chamber and organic fuel injection means to burn with some of the organic fuel on the outside of the combustion chamber, and thus be in direct heat exchange relation with the organic fuel going into the catalyst bed.

Vanderborgh, Nicholas E. (Los Alamos, NM); Springer, Thomas E. (Los Alamos, NM); Huff, James R. (Los Alamos, NM)

1987-01-01

160

Nuclear fuel element assemblies  

International Nuclear Information System (INIS)

A fuel element assembly for a high temperature reactor comprises a prismatic block having fuel containing bores and interstitial coolant conducting bores extending end-to-end. The fuel comprises stacks of annular compacts which line the fuel containing bores and define central coolant flow channels through the fuel. (U.S.)

161

Fuel Cell Overview  

Science.gov (United States)

This presentation from Project Lead the Way Ohio looks at fuel cells. The origins of the technology, how fuel cells work and modern applications of fuel cell technologies are discussed. Information on different types of fuel cells and their potential use in fueling automobiles is also included. This document may be downloaded in Microsoft PowerPoint file format.

162

Instrumentation of fuel elements and fuel plates  

International Nuclear Information System (INIS)

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)

163

KMRR fuel design  

International Nuclear Information System (INIS)

KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

164

Fuel behaviour and fuel pin failure probability  

International Nuclear Information System (INIS)

Owing to anticipated detrimental effects of low fuel sintering density with respect to the combined mechanical/chemical interactions of fuel and cladding, far higher fuel density appears to be imperative. There has been no doubt about the aptidude of the reference cladding steel-DIN-I-4970 under the aspects mentioned herein, whereas the cladding steel-DIN-I-4981 is likely to show weaknesses concerning the combined mechanical/chemical interactions with the fuel. The possibility of peaks in cladding strain caused by an accumulation of cesium needs checking out with standard-length fuel pins. After a long period of reactor power derating or predictable raise of rod power, a power increase is to be done very slowly. For fuel pins having a low fuel sintering density low rod power operation at the start does not seem to be advantageous. (orig./RW)

165

Reactor fuels and materials  

Energy Technology Data Exchange (ETDEWEB)

This book deals with reactor fuels and materials, introducing elementary knowledge of materials such as interaction between radiation and crystal lattice, damage of the crystal by the radiation and recovery, uranium enrichment and a nuclear fuel reprocessing, metal fuels, uranium silicide fuels, covering material fuels, uranium dioxide fuels, gas cooled reactor materials fast reactor materials, fusion of cell nuclear, control materials of neutron and moderator materials. This books explains how to deal with these materials and how to use them in right way.

Lee, Gi Sun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-08-15

166

Fuel performance annual report  

International Nuclear Information System (INIS)

This article summarizes NUREG/CR-3950 (PNL-5210), vols. 1 and 2, entitled Fuel Performance Annual Report for 1983 and Fuel Performance Annual Report for 1984, respectively. These reports, the sixth and seventh in a series, were published in March 1985 and March 1986, respectively, to provide a summary description of fuel performance in commercial nuclear power plants during 1983 and 1984. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high burnup fuel experience, and items of general significance are provided. References to additional, more detailed information and related Nuclear Regulatory Commission evaluations are included

167

Fuel performance experience  

International Nuclear Information System (INIS)

The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

168

Nuclear fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To enable ensure reactor shut-down margin required for the reactor shut-down even when the enrichment degree of fuel material is increased in a BWR type reactor. Constitution: A low nuclear fuel charging rate of the fissionable or fertile material per unit length in the height direction of nuclear fuel assembly is set to be lower in the upper portion than the other portion of the nuclear fuel assembly. Hollow fuel pellets each having a hollow portion at the center or fuel pellets fabricated to lower density than that of other fuel pellets are used in the low nuclear fuel charging rate region. (Seki, T.)

169

HTGR fuel reprocessing technology  

International Nuclear Information System (INIS)

The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

170

Fuel cells: principles, types, fuels, and applications.  

Science.gov (United States)

During the last decade, fuel cells have received enormous attention from research institutions and companies as novel electrical energy conversion systems. In the near future, they will see application in automotive propulsion, distributed power generation, and in low power portable devices (battery replacement). This review gives an introduction into the fundamentals and applications of fuel cells: Firstly, the environmental and social factors promoting fuel cell development are discussed, with an emphasis on the advantages of fuel cells compared to the conventional techniques. Then, the main reactions, which are responsible for the conversion of chemical into electrical energy in fuel cells, are given and the thermodynamic and kinetic fundamentals are stated. The theoretical and real efficiencies of fuel cells are also compared to that of internal combustion engines. Next, the different types of fuel cells and their main components are explained and the related material issues are presented. A section is devoted to fuel generation and storage, which is of paramount importance for the practical aspects of fuel cell use. Finally, attention is given to the integration of the fuel cells into complete systems. PMID:23696319

Carrette, L; Friedrich, K A; Stimming, U

2000-12-15

171

Spent fuel consolidation system  

International Nuclear Information System (INIS)

The spent fuel consolidation system provides method and apparatus for remotely vertically and horizontally compacting an array of spent fuel rods while the fuel rods remain submerged in a coolant. The invention comprises a row ordering section for rearranging the configuration of the fuel rods, horizontal consolidation section for horizontally compacting several rows of fuel rods, and a vertical consolidation section for vertically compacting several rows of horizontally compacted fuel rods. The system is capable of compacting the fuel rods from a given fuel assembly to about one half of the volume originally occupied by such fuel rods in the fuel assembly thereby providing greater storage capacity for a given volume of spent fuel storage

172

Nuclear fuel cycle  

International Nuclear Information System (INIS)

Problems of the full fuel cycle, like uranium exploration, enrichment, conversion into nuclear fuel and spent fuel management are discussed. Forecasting uranium supply and demand until 2010 and factors affecting the uranium market are considered. Partitioning/transmutation and spent fuel management programs of the IAEA are highlighted

173

Fuel Cells Fact Sheet  

Science.gov (United States)

This document provides a basic introduction to fuel cells: how they work, the different types of fuel cells (PEM, AFC, PAFC, DMFC, MCFC and SOFC) and the advantages and disadvantages of using fuel cells. Two useful graphic representations of fuel cells are also included. This document may be downloaded in PDF file format.

174

Fuel element design handbook  

Energy Technology Data Exchange (ETDEWEB)

The economic development of nuclear reactors depends upon the integrated progress in the fields of reactor design, fuel element design, reactor operation, and fuel production and separation. Broad criteria, which restrict the fuel element design, are determined by the mutual consideration of the problems encountered in all the above fields. Hence, no stage of reactor design or operation is independent of the fuel element problem, nor can the fuel element designer disregard the interest of any one field. As an introduction to the fuel element design problem, this chapter describes how the general criteria for a fuel element are determined.

Merckx, K.R.

1958-09-01

175

Nuclear fuel elements  

International Nuclear Information System (INIS)

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

176

PHWR fuel fabrication  

International Nuclear Information System (INIS)

With the decision of the Indian Department of Atomic Energy to opt for a Heavy Water Reactor system for nuclear power generation work was taken up in the early sixties on development of technology on Zircaloy-clad U02 fuels. Pilot scale production facilities were set up at Trombay to evolve the technology of U02 powder production and fuel element fabrication. Half the initial charge of fuel was made in India and the other half was provided by Canada. The fuel Zircaloy-clad natural U02 was made with imported zircaloy tubes and hardware for the first half charge fuel. The experience thus gained was used to design and build a large-scale fuel fabrication facility. The Nuclear Fuel Complex (NFC) thus came into existence in early 70's to manufacture PHWR fuel. Facilities were also established at NFC for production of BWR fuel, zircaloy tubing and hardware required for the fuel; and Zircaloy coolant and calandria tubes required for the reactors. NFC produces Zircaloy-clad natural U02 fuel for PHWRs at Kota (Rajasthan), Kalpakkam (Tamil Nadu) and Narora (Uttar Pradesh) starting from indigenous magnesium diuranate concentrates from (Uranium Corporation of India Limited ) for production of U02 pellets, and zircon beach sands from IRE (Indian Rare Earths Limited ) for production of Zircaloy fuel and hardware. The fuel production plants are being expanded to meet the increased fuel requirements of the planned nuclear power pequirements of the planned nuclear power programme. The fuel produced so far has shown an excellent in-reactor behaviour as judged by the very low failure rates. With the development of computer codes for fuel design and management and with the establishment of fuel design and testing capabilites, 'total fuel' capability has been established leading to self-sufficiency in this vital area of nuclear technology. This paper primarily details our experience in fuel manufacture and inspection and highlights operational experience

177

Production and characterization of atomized U-Mo powder by the rotating electrode process  

International Nuclear Information System (INIS)

In order to produce feedstock fuel powder for irradiation testing, the Idaho National Laboratory has produced a rotating electrode type atomizer to fabricate uranium-molybdenum alloy fuel. Operating with the appropriate parameters, this laboratory-scale atomizer produces fuel in the desired size range for the RERTR dispersion experiments. Analysis of the powder shows a homogenous, rapidly solidified microstructure with fine equiaxed grains. This powder has been used to produce irradiation experiments to further test adjusted matrix U-Mo dispersion fuel. (author)

178

BWR fuel performance  

International Nuclear Information System (INIS)

The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

179

Nuclear fuel element  

International Nuclear Information System (INIS)

A nuclear fuel element is constituted by incorporating nuclear fuel pellets in a state that a plurality of the fuel pellets are stacked in a fuel cladding tube having an inner surface integrally lined with a liner layer. The nuclear fuel pellet is formed by compressing uranium dioxide powder into a cylindrical shape followed by sintering, and the end face is chamfered. The liner layer comprises zirconium containing iron and oxygen as main impurities. The fuel cladding tube comprises zircaloy-2. The size of the space caused by chamfering the end surface of the nuclear fuel pellet and the liner layer are determined such that the maximum length in the longitudinal direction of the space caused by chamfering the end face of the nuclear fuel pellet is not greater than twice of the thickness of the liner. This can eliminate fine perforating cracks on the inner surface of the lined fuel cladding tube. (I.N.)

180

DUPIC fuel compatibility assessment  

International Nuclear Information System (INIS)

The purpose of this study is to assess the compatibility of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The Phase II study of this project includes the analysis of impact on the reactor safety, the development of core design technology, the development of fuel supply technology of optimal composition, and feasibility analysis on localization and license of DUPIC fuel. From the reactor safety analysis results, it is known that DUPIC fuel satisfies the safety limit of reactor containment and public dose for single failure. But, the safety limit may be exceeded for dual failure. Therefore, more analysis is needed for the removal of excessive conservatism in accident analysis methodology and modification of transient fuel behavior analysis methodology. The results of the validation calculations of core design methodology have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of compatibility and fuel fabrication have shown that DUPIC fuel is technically feasible. For practical use and licensing, however, more research items required in the practical use, fuel rod and bundle design and fuel loading are should be performed. When these items are performed and resolved, the compatibility of the DUPIC fuel is achieved, and, eventually, the possibility of DUPIC fuel licensing can be confirmed

181

Micro fuel cell  

Energy Technology Data Exchange (ETDEWEB)

An ambient temperature, liquid feed, direct methanol fuel cell device is under development. A metal barrier layer was used to block methanol crossover from the anode to the cathode side while still allowing for the transport of protons from the anode to the cathode. A direct methanol fuel cell (DMFC) is an electrochemical engine that converts chemical energy into clean electrical power by the direct oxidation of methanol at the fuel cell anode. This direct use of a liquid fuel eliminates the need for a reformer to convert the fuel to hydrogen before it is fed into the fuel cell.

Zook, L.A.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States); Hockaday, R. [Energy Related Devices Inc., Los Alamos, NM (United States)

1998-12-31

182

Nuclear fuel storage  

International Nuclear Information System (INIS)

A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

183

Fuel inspection techniques  

International Nuclear Information System (INIS)

Poolside fuel inspections are performed to measure fuel performance characteristics such as corrosion, rod growth, assembly growth, rod bow, and cladding creep. These inspections address the performance of intact fuel assemblies and individual removed fuel rods. Poolside inspections are also used to identify the cause of leaking fuel rods once they have been identified as leaking by sipping, ultrasonic or eddy current testing. Various inspections, such as fuel rod visuals, fiberscope of grid cells, profilometry, and cell sizing can be performed to identify the root cause of leaking rods. These examinations provide high quality data safely and reliably with minimal plant impact. (author)

184

Multiple fuel rod gripper  

International Nuclear Information System (INIS)

The multiple fuel rod gripper comprises a plurality of split tube collets arranged to be inserted into corresponding tapered holes in a locking plate. When the gripper has been positioned to have a plurality of fuel rods disposed in the holes of locking plate, an actuating mechanism causes the collets to be inserted into the tapered holes thereby causing the fuel rods to be inserted in the collets. The taper of the holes forces the collets into locking engagement with the fuel rods so that the fuel rods may be extracted from the fuel assembly

185

DUPIC fuel compatibility assessment  

International Nuclear Information System (INIS)

The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

186

Electronuclear fissile fuel production  

International Nuclear Information System (INIS)

A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for three 1000 MW(e) LWR power reactors overs its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center. (orig.)

187

Romanian nuclear fuel program  

International Nuclear Information System (INIS)

The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle of January 1997 with fuel produced by the Romanian fuel plant. The quality evaluation of the 'pre-1990' fuel started in April 1996 and was performed by the Nuclear Fuel Plant (FCN) Pitesti, under the supervision of the Nuclear Power Group (GEN) - a distinct department of RENEL. The paper presents the involvement of Romania in the activities related to the Advanced CANDU Fuel Cycle. The future prospect and trend of the Romanian Nuclear Fuel Program are also presented in this paper. (author)

188

Fuel Assembly Damping Summary  

International Nuclear Information System (INIS)

This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping measurement testing under flow are also briefly discussed. Fuel assembly damping is an essential parameter to determine fuel assembly dynamic behavior in operating or accidental core. Dry damping coefficient from the out-pile pluck testing was used for the accident analysis model in conservative and simplified manner. But, this is way lower than wet or under-flow damping

189

Fuel Assembly Damping Summary  

Energy Technology Data Exchange (ETDEWEB)

This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping measurement testing under flow are also briefly discussed. Fuel assembly damping is an essential parameter to determine fuel assembly dynamic behavior in operating or accidental core. Dry damping coefficient from the out-pile pluck testing was used for the accident analysis model in conservative and simplified manner. But, this is way lower than wet or under-flow damping.

Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

190

EOS Reactor Fuel  

International Science & Technology Center (ISTC)

Development of the Equations of State for Fuel Compositions, which Take into Account the Microstructure Accumulation Kinetics its Use for Simulation of the Fuel Failure Consequences in Nuclear Reactors of VariousType. (Continuation of the project 003)

191

Nuclear fuel element  

International Nuclear Information System (INIS)

The present invention concerns a nuclear fuel element for an LMFBR type reactor. Cadmium (Cd) is previously deposited and uniformly dispersed on the inner circumferential surface of a fuel can made of a zirconium (Zr) based alloy. With such a constitution, localized Cd nuclei are formed and there can be expected for a getter effect of Cd released from UO2 during use of fuels. That is, Cd vapors as corrosive fission products formed on the inner surface of the fuel can are adsorbed on and react with Cd-Zr products formed previously on the inner circumferential surface of the fuel can. Accordingly, since the possibility of Cd vapors to reach the top end of crack in the fuel can, the brittle cracks of the fuel can are less induced. Accordingly, safety and economy of nuclear fuel elements can be improved. (I.N.)

192

Fuel cycle data survey  

International Nuclear Information System (INIS)

A survey of the fuel cycle cost data published during 1977 and 1978 is presented in tabular and graphical form. Cost trends for the period 1965 onwards are presented for yellow cake, conversion, uranium enrichment, fuel fabrication and reprocessing

193

Fuel assembly guide tube  

International Nuclear Information System (INIS)

This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

194

Future automotive fuels  

International Nuclear Information System (INIS)

There are several important factors which are fundamental to the choice of alternative automobile fuels: the chain of energetic efficiency of fuels; costs; environmental friendliness; suitability for usual engines or adapting easiness; existing reserves of crude oil, natural gas or the fossil energy sources; and, alternatively, agricultural potentiality. This paper covers all these factors. The fuels dealt with in this paper are alcohol, vegetable oil, gaseous fuel, hydrogen and ammonia fuels. Renewable fuels are the most valuable forms of renewable energy. In addition to that rank, they can contribute to three other problem areas: agricultural surpluses, environmental degradation, and conservation of natural resources. Due to the competitive utilization of biomass for food energy production, bio-fuels should mainly be produced in those countries where an energy shortage is combined with a food surplus. The fuels arousing the most interest are alcohol and vegetable oil, the latter for diesel engines, even in northern countries. (au)

195

The nuclear fuel cycle  

International Nuclear Information System (INIS)

This chapter explains the distinction between fissile and fertile materials, examines briefly the processes involved in fuel manufacture and management, describes the alternative nuclear fuel cycles and considers their advantages and disadvantages. Fuel management is usually divided into three stages; the front end stage of production and fabrication, the back end stage which deals with the fuel after it is removed from the reactor (including reprocessing and waste treatment) and the stage in between when the fuel is actually in the reactor. These stages are illustrated and explained in detail. The plutonium fuel cycle and thorium-uranium-233 fuel cycle are explained. The differences between fuels for thermal reactors and fast reactors are explained. (U.K.)

196

Fuel Cell Applications  

Science.gov (United States)

This page uses flash animation to briefly explain the many areas where fuel cell technology can be applied. It also discusses the need for alternative energies as well as outlines the advantages of fuel cells.

197

Nuclear fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To prevent failures in fuel cans by decreasing the amount of abrasion due to fretting corrosion in the outer surface of the fuel cans for fuel elements. Constitution: Film of abrasion-resistant material is formed to the outer surface to fuel cans or on the surface of the fuel cans and the spacer surface in a fuel rod held by spacers. The film-forming material is, desirably, composed of an abrasion-resistant substance of a low friction coefficient such as graphite and chromium. The film made of such material is formed by way of electroplating, chemical plating, vapor deposition, electrodeposition and the like into a thickness, preferably of above 5 microns. The membrane thus formed can provide a high abrasion resistance upon mounting the spacers to the fuel rod and prevent fretting corrosion to thereby avoid failures in the fuel cans. (Moriyama, K.)

198

Nuclear fuel element  

Science.gov (United States)

A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

Zocher, Roy W. (Los Alamos, NM)

1991-01-01

199

Reformulated diesel fuel  

Science.gov (United States)

Reformulated diesel fuels for automotive diesel engines which meet the requirements of ASTM 975-02 and provide significantly reduced emissions of nitrogen oxides (NO.sub.x) and particulate matter (PM) relative to commercially available diesel fuels.

McAdams, Hiramie T [Carrollton, IL; Crawford, Robert W [Tucson, AZ; Hadder, Gerald R [Oak Ridge, TN; McNutt, Barry D [Arlington, VA

2006-03-28

200

Fuel element assembly  

International Nuclear Information System (INIS)

This invention relates to fuel element assemblies for ?tight lattice? water-cooled nuclear converter reactors in which fission is induced predominantly by neutrons with energy levels beyond the range of the thermal neutron spectrum. The assembly provides for a multiplicity of cylindrical fuel rods arranged parallel to each other in a spaced array having an equilateral-triangular pattern. External longitudinal fins serially located at space intervals along the length of each fuel rod curve about a portion of the fuel rod at a constant angle, and tangentially contact the surface of an adjacent fuel rod thereby effecting a mutual six-point lateral support. The fins are fixed to the adjacent fuel elements at the points of tangential contact in the same lateral plane such that the fuel elements and fins comprise a single unit within the fuel assembly

201

Direct hydrocarbon fuel cells  

Science.gov (United States)

The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

Barnett, Scott A.; Lai, Tammy; Liu, Jiang

2010-05-04

202

Control of Fuel Cells  

OpenAIRE

This thesis deals with control of fuel cells, focusing on high-temperature proton-exchange-membrane fuel cells. Fuel cells are devices that convert the chemical energy of hydrogen, methanol or other chemical compounds directly into electricity, without combustion or thermal cycles. They are efficient, scalable and silent devices that can provide power to a wide variety of utilities, from portable electronics to vehicles, to nation-wide electric grids. Whereas studies about the design of fuel ...

Zenith, Federico

2007-01-01

203

Fuel cell electronics packaging  

CERN Document Server

Today's commercial, medical and military electronics are becoming smaller and smaller. At the same time these devices demand more power and currently this power requirement is met almost exclusively by battery power. This book includes coverage of ceramic hybrid separators for micro fuel cells and miniature fuel cells built with LTCC technology. It also covers novel fuel cells and discusses the application of fuel cell in microelectronics.

Kuang, Ken

2007-01-01

204

Solid Oxide Fuel Cells  

Science.gov (United States)

This reference sheet provides some basic information on solid oxide fuel cells. This document includes information on the basic operation of these fuel cells and some useful graphics. This document would probably be more useful for students who already have a basic understanding of fuel cells.This document may be downloaded in PDF file format.

205

Nuclear fuel cycle  

International Nuclear Information System (INIS)

Status of different nuclear fuel cycle phases in 1992 is discussed including the following issues: uranium exploration, resources, supply and demand, production, market prices, conversion, enrichment; reactor fuel technology; spent fuel management, as well as trends of these phases development up to the year 2010. 10 refs, 11 figs, 15 tabs

206

Denatured fuel cycles  

International Nuclear Information System (INIS)

This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

207

Clickable Fuel Cell Car  

Science.gov (United States)

In this interactive, students can investigate a typical hydrogen fuel cell prototype car from its fuel cell stacks to its ultracapacitor, a kind of supplementary power source. The limited-production vehicle seen in this feature is a Honda 2005 FCX, which is typical of the kinds of hydrogen fuel cell cars that some major automakers are now researching and developing.

Peter Tyson

208

Modeling: driving fuel cells  

Directory of Open Access Journals (Sweden)

Fuel cells were invented in 1839 by Sir William Grove, a Welsh judge and gentleman scientist, as a result of his experiments on the electrolysis of water. To put it simply, fuel cells are electrochemical devices that take hydrogen gas from fuel, combine it with oxygen from the air, and generate electricity and heat, with water as the only by-product.

Michael Francis

2002-05-01

209

DUPIC fuel compatibility assessment  

International Nuclear Information System (INIS)

In this report, analysis results for the CANDU 6 reactor with DUPIC fuel have been described. Various problems are assessed against the standard natural uranium fuel core such as fuel fabrication, fuel rod and bundle design, in-core loading, in-core fuel management, spent fuel treatment and overall fuel cycle. Some of the results are related to the license and demonstration. From the up to date results, it is known that the DUPIC fuel fabrication is technically feasible and the anticipated in-core problems can be resolved by current technique. Also, the benefit is expected in power distribution and fuel burnup. However, because the CANDU 6 reactor is originally designed for natural uranium fuel, some demerits are found in some field such as radiation damage of the reactor structural material, operational margin decrease by composition heterogeneity, increase in fission product release of accident condition, deterioration of fuel pellet material property. These problems should be resolved technically including design improvement of DUPIC fuel and CANDU 6 reactor. Furthermore, experimental verifications should be performed for reactor physics and thermal hydraulics. This report describes the compatibility with the CANDU 6 reactor, and it should be noted that detail and wide work should be performed for more reliable results

210

Westinghouse fuel pellet evolution  

International Nuclear Information System (INIS)

Recognizing fuel reliability, fuel cycle cost and security of supply as key customer expectations, Westinghouse has developed a comprehensive strategy for fuel pellet evolution. It encompasses state of art flawless manufacturing and a superior irradiation behavior as well as standardization across manufacturing facilities

211

Cracked fuel mechanics  

International Nuclear Information System (INIS)

Fuel pellets undergo thermally induced cracking during normal reactor operation. Some fuel performance codes have included models that address the effects of fuel cracking on fuel rod thermal and mechanical behavior. However, models that rely too heavily on continuum mechanics formulations (annular gaps and solid cylindrical pellets) characteristically do not adequately predict cladding axial elongations. Calculations of bamboo ridging generally require many assumptions concerning fuel geometry, and some of the methods used are too complex and expensive to employ on a routine basis. Some of these difficulties originate from a lack of definition of suitable parameters which describe the cracked fuel medium. The methodology is being improved by models that describe cracked fuel behavior utilizing parameters with stronger physical foundations instead of classical continuum formulations. This paper presents a modelling concept and a set of measurable parameters that have been shown to improve the prediction of the mechanical behavior of cracked fuel/cladding systems without added computational expense. The transition from classical annular gap/cylindrical pellet models to modified bulk properties and further to local behavior for cracked fuel systems is discussed. The results of laboratory experiments to verify these modelling parameters are shown. Data are also presented from laboratory experiments on unirradiated and irradiated rods which show that fuel rod mechanical response depends on fuel fragment size. The impact of these data on cracked fuel behavior and failure modelling is also discussed. (author)

212

Canadian Renewable Fuels Association  

Science.gov (United States)

The CRFA promotes the use of renewable bio-fuels (ethanol, biodiesel). Membership includes representatives from fuel marketing, fuel production agriculture, forestry, engineering and environmental organizations, and researchers and individuals. Visitors can find policy papers, industry statistics, plant locations, and FAQs and fact sheets about biodiesel and ethanol.

213

Fuel Cell Technologies Program  

Science.gov (United States)

This document from the U.S. Department of Energy provides an introduction to fuel cell technology. The material outlines how they work, and why they may be chosen as a fuel source. Different types of fuel cells, their applications, advantages and disadvantages are outlined. This document may be downloaded in PDF file format.

214

Direct Methanol Fuel Cell  

Science.gov (United States)

This sheet provides information about direct methanol fuel cells. Details on the chemistry involved are included in graphic form along with several notes on these fuel cells. This material would be most appropriate for upper level students who already have a basic understanding of fuel cell technology and chemistry. This document may be downloaded in PDF file format.

215

Molten Carbonate Fuel Cell  

Science.gov (United States)

This page is an introduction to the Molten Carbonate fuel cell. It uses flash animation to explain in greater detail what the Molten Carbonate fuel cell consists of and how it works. The website has an introductory animation which is followed by more in depth description of the molten carbonate fuel cell works.

216

Solid Oxide Fuel Cell  

Science.gov (United States)

This page is an introduction to the Solid Oxide fuel cell. It uses flash software to explain in greater detail what the Solid Oxide fuel cell consists of and how it works. The website has an introductory animation which is followed by more in depth description of the solid oxide fuel cell.

217

Hydrogen and fuel cells  

International Nuclear Information System (INIS)

This road-map proposes by the Group Total aims to inform the public on the hydrogen and fuel cells. It presents the hydrogen technology from the production to the distribution and storage, the issues as motor fuel and fuel cells, the challenge for vehicles applications and the Total commitments in the domain. (A.L.B.)

218

Korean nuclear fuel program  

International Nuclear Information System (INIS)

In Korea, sixteen PWRs and four PHWRs are in operation, and eight more PWRs are to be built by the year 2015. Sixteen operating PWRs comprise of eight Westinghouse type plants and eight Optimized Power Reactors. Korea Nuclear Fuel Company is designing, manufacturing and supplying nuclear fuel for all operating PWRs and PHWRs in Korea since 1989. It has performed reactor core design and safety analysis for more than 16 initial cores and 130 reload cores as of the end of 2004. It has been manufacturing PWR fuel with the annual production of 400 ton-U as well as PHWR fuel with the annual production of 400 ton-U. Korea Nuclear Fuel Company launched a 'Mid- and Long-term Nuclear Fuel Development Program' in 1999. It has completed the development of PLUS7TM fuel in 2002, which is an advanced fuel for the fuel for the Optimized Power Reactors and the Advanced Power Reactors. Region application of PLUS7TM is scheduled to begin in 2006. It has also completed the development of both 16ACE7TM and 17ACE7TM fuel in 2004, which are advanced fuel for the 2-loop and 3-loop Westinghouse type plants. Region Application of 16ACE7TM and 17ACE7TM is planed to start in 2008 and in 2009, respectively. To be more competitive than ever and to become one of global fuel technology leaders, KNFC launched 'Green Vision 2015 Projects' in the year 2005. (author)

219

Barrier fuel counters pci  

International Nuclear Information System (INIS)

A new 'barrier' fuel developed by General Electric for boiling water reactors is described. A thin protective layer of soft zirconium absorbs a fuel pellet's expansion and inhibits chemical attack. This design is expected to counter the pellet-clad-interaction mechanism, which will mean even higher levels of fuel reliability and improved plant operating flexibility. (U.K.)

220

AFIP-4 Irradiation Summary Report  

Energy Technology Data Exchange (ETDEWEB)

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

2011-09-01

221

AFIP-4 Irradiation Summary Report  

Energy Technology Data Exchange (ETDEWEB)

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

2012-01-01

222

AFIP-6 Irradiation Summary Report  

Energy Technology Data Exchange (ETDEWEB)

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-6 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a length prototypic to that of the ATR fuel plates (45 inches in length). The AFIP-6 test was the first test with plates in a swaged condition with longer fuel zones of approximately 22.5 inches in length1,2. The following report summarizes the life of the AFIP-6 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Danielle M Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

2011-09-01

223

Oconee spent fuel rerack  

International Nuclear Information System (INIS)

Spent fuel storage problems facing electric utilities with nuclear generation are growing more critical as existing spent fuel storage capacity is utilized. Due to the inaccessibility of spent fuel reprocessing plants, alternative temporary solutions such as transfer of spent nuclear fuel to other storage facilities and increasing the capacity of existing storage facilities through reracking are becoming increasingly prevalent. This paper describes the method and installation of new racks for increasing the fuel storage capacity of unit 3 of Duke Power Company's Oconee Nuclear Station near Seneca, South Carolina

224

Plutonium fuel program  

International Nuclear Information System (INIS)

The work of the project Fuel Development in 1976 was marked by three important developments. Firstly, the reproduceability of the process to produce sphere pac carbide fuel by a gelation process was established. Secondly, in the post irradiation examination of the fuel pins from the BR-2 reactor, the fuel reached approximately 5.5% FIMA without failure. Thirdly, outside interest in sphere pac material became more apparent. These developments are discussed, and plans to construct a fuel pilot plant to go into operation in the 1980's are revealed. (Auth.)

225

Nuclear fuel cycle costs  

International Nuclear Information System (INIS)

The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

226

Cracked-fuel mechanics  

International Nuclear Information System (INIS)

This paper presents a modelling concept and a set of measurable parameters that have been shown to improve the prediction of the mechanical behavior of cracked fuel/cladding systems without added computational expense. The transition from classical annular gap/cylindrical pellet models to modified bulk properties and further to local behavior for cracked fuel systems is discussed. The results of laboratory experiments to verify these modelling parameters are shown. Data are also presented from laboratory experiments on unirradiated and irradiated rods which show that fuel rod mechanical response depends on fuel fragment size. The impact of these data on cracked fuel behavior and failure modelling is also discussed

227

Mox fuels recycling  

International Nuclear Information System (INIS)

This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

228

Mox fuels recycling  

Energy Technology Data Exchange (ETDEWEB)

This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

Gay, A. [Cogema, 78 - Velizy Villacoublay (France)

1998-07-01

229

Nuclear reactor fuel element  

International Nuclear Information System (INIS)

The invention concerns fuel elemnts, where the guide tubes for the control rods are made of zirconium alloys (e.g. Zircaloy), while the spacing grid, the fuel element top and the fuel element bottom consist of steel. In order to connect these different materials together reliably, it is proposed to provide envelopes of material which is the same as that of the fuel element parts, which are metallurgically connected with parts made of the same material, while they are secured to other parts of the fuel elements by positive mechanical locking. The subclaims explain preferred constructional details by means of 13 drawings. (UWI) 891 HP/UWI 892 MB

230

Fuel Cell Today  

Science.gov (United States)

Fuel Cells Today is a useful online resource with a very diverse range of materials about fuel cell technology. Possibly the most interesting part of the site is the Reference Centre, where users can find information on different types of fuel cells, their applications, history of their development, possible materials to use in their design, and more. All educational and technical descriptions are intended to promote the global adoption of fuel cells as a clean, efficient energy source. There is also plenty of literature in the Knowledge Bank. Fuel cell news and emerging technologies are covered, and the site is updated often.

231

Engine fuels from biomass  

Science.gov (United States)

Sources of biomass fuels for engines are compared to other synfuels. Biomass can be converted to gaseous and liquid engine fuels by the same processes utilized for coal conversion such as gasification, direct liquefaction, and indirect liquefaction. Alternatively, biomass can be converted into liquid fuels by fermentation to methane or ethanol. The quantities of biomass derived engine fuels potentially available in the next decade are relatively small, and the anticipated costs are significantly greater than for liquid engine fuels made from coal or oil shale.

Parker, H. W.

1981-01-01

232

Inspection system for fuels  

International Nuclear Information System (INIS)

A typical embodiment of the invention combines a novel cellular end fitting for a nuclear reactor fuel assembly with a new design for a fuel rod end cap and radiation sensing device probe to provide a means for swiftly and accurately distinguishing sound fuel rods from those rods that have developed leaks. For example, a somewhat thinner than usual fuel rod end cap is accessible through the open cellular structure of the end fitting to permit a hollow metal probe to contact the fuel rod end cap. This direct contact excludes most of the water, metal and other shielding materials from the volume between the interior of the fuel and the radiation detector, thereby improving the quality of the fuel rod examination. A bridge and trolley structure for accurately positioning the probe also is described. (Auth.)

233

Fuel pin bundle splitting  

International Nuclear Information System (INIS)

The patent describes the splitting of a bundle of nuclear fuel pins into smaller bundles, during the dismantling of a fuel element, in preparation for the reprocessing of the spent fuel. The size of the small bundles are such that they are suitable for cropping in an easily maintainable shearing machine. The cropping of fuel pins into short sections exposes the irradiated fuel to be reprocessed. The invention involves feeding a number of blades into the exposed end of a fuel pin bundle. The bundle is forced out of the containing sheath by a ram, and the fuel pins are forced to pass either side of theblades, there by the bundle is sorted into a number of smaller bundles. (U.K.)

234

Nuclear fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To effectively cool a fuel even at the time of loss-of-primary coolant accident. Constitution: More than 35% of total fuel rods contained in a fuel assembly are projected at their upper ends above an upper tie plate. Since the diffusion of the air blown up into a space is suppressed due to the presence of the projection in this case and the air diffused to the upper portion of a through hole is thus reduced, water heat pressure becomes large proportionally thereto so that the quantity of the dropped water increases. Accordingly, even if not all the fuel rods are projected at their upper ends but partial fuel rods are merely projected, it can exert the effect of diffusing the air in the space. In this manner, since the quantity of sprayed flow rate into the fuel assembly is increased, the temperature of the fuel cladding tube at the time of failure can be reduced. (Yoshihara, H.)

235

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To improve the reliablity of fuel elements by forming recesses to the outer circumference of nuclear fuel pellets to thereby prevent stress corrosion cracks in the fuel cans due to thermal stress cracks in the pellet. Constitution: A plurality of recesses or annular grooves are formed generally along the circular circumference on the outer periphery of a sintered fuel pellets axially stacked in plurality and sealed in a fuel can. The shoulder of the above recesses or the annular groove is subjected to grinding fabrication. The presence of the ground surface forms surface contaction between the pellet and the can even if cracks are resulted in the pellets due to thermal stress during reactor operation thereby reducing the local stress applied to the fuel can, whereby the stress corrosion cracks can be prevented and the reliability for the fuel elements can be improved. (Moriyama, K.)

236

Fuel nozzle assembly  

Science.gov (United States)

A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

Johnson, Thomas Edward (Greer, SC); Ziminsky, Willy Steve (Simpsonville, SC); Lacey, Benjamin Paul (Greer, SC); York, William David (Greer, SC); Stevenson, Christian Xavier (Inman, SC)

2011-08-30

237

Fuel Cells 2000  

Science.gov (United States)

Fuel Cells 2000, an organization dedicated to informing the public about fuel cells, offers this website with an interactive map listing companies and research organizations connected with the U.S. fuel cell industry. A second map shows U.S. Fuel Cell Installations and Vehicle Demonstrations. Links to the organizations' websites make this an easy-to-use resource for finding out more about fuel cells and looking up local demonstrations. Visitors can also download a full directory of nearly 1000 fuel-cell related companies and organizations and a chart showing fuel cell installations worldwide. (Unfortunately, many of the other links on this website were not working at the time of this writing.)

238

Nuclear fuel cycles  

International Nuclear Information System (INIS)

The source of energy in the nuclear reactors in fission if a heavy nuclei by absorbing a neutron and giving fission products, few neutrons and gamma radiation. The Nuclear Fuel Cycle may be broadly defined as the set of process and operations needed to manufacture nuclear fuels, to irradiate them in nuclear reactors and to treat and store them, temporarily or permanently, after irradiation. Several nuclear fuel cycles may be considered, depending on the type of reactor and the type of fuel used and whether or not the irradiated fuel will be reprocessed. The nuclear fuel cycle starts with uranium exploration and ends with final disposal of the material used and generated during the cycle. For practical reasons the process has been further subdivided into the front-end and the back-end. The front-end of the cycle occurs before irradiation and the back-end begins with the discharge of spent fuel from the reactor

239

Oxy-fuel combustion of solid fuels  

Energy Technology Data Exchange (ETDEWEB)

Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO{sub 2} from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame temperature. The flue gas produced thus consists primarily of carbon dioxide and water. Much research on the different aspects of an oxy-fuel power plant has been performed during the last decade. Focus has mainly been on retrofits of existing pulverized-coal-fired power plant units. Green-field plants which provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy-fuel process and focuses particularly on the combustion fundamentals, i.e. flame temperatures and heat transfer, ignition and burnout, emissions, and fly ash characteristics. Knowledge is currently available regarding both an entire oxy-fuel power plant and the combustion fundamentals. However, several questions remain unanswered and more research and pilot plant testing of heat transfer profiles, emission levels, the optimum oxygen excess and inlet oxygen concentration levels, high and low-temperature fire-side corrosion, ash quality, plant operability, and models to predict NO{sub x} and SO{sub 3} formation is required. (author)

Toftegaard, Maja B. [Department of Chemical and Biochemical Engineering, Technical University of Denmark, DK-2800 Kgs. Lyngby (Denmark); DONG Energy, Kraftvaerksvej 53, DK-7000 Fredericia (Denmark); Brix, Jacob; Jensen, Peter A.; Glarborg, Peter; Jensen, Anker D. [Department of Chemical and Biochemical Engineering, Technical University of Denmark, DK-2800 Kgs. Lyngby (Denmark)

2010-10-15

240

Oxy-fuel combustion of solid fuels  

Energy Technology Data Exchange (ETDEWEB)

Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO{sub 2} from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame temperature. The flue gas produced thus consists primarily of carbon dioxide and water. Much research on the different aspects of an oxy-fuel power plant has been performed during the last decade. Focus has mainly been on retrofits of existing pulverized-coal-fired power plant units. Green-field plants which provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy-fuel process and focuses particularly on the combustion fundamentals, i.e. flame temperatures and heat transfer, ignition and burnout, emissions, and fly ash characteristics. Knowledge is currently available regarding both an entire oxy-fuel power plant and the combustion fundamentals. However, several questions remain unanswered and more research and pilot plant testing of heat transfer profiles, emission levels, the optimum oxygen excess and inlet oxygen concentration levels, high and low-temperature fire-side corrosion, ash quality, plant operability, and models to predict NOx and SO{sub 3} formation is required. 218 refs., 48 figs., 13 tabs.

Maja B. Toftegaard; Jacob Brix; Peter A. Jensen; Peter Glarborg; Anker D. Jensen [Technical University of Denmark, Lyngby (Denmark). Department of Chemical and Biochemical Engineering

2010-10-15

241

Oxy-fuel combustion of solid fuels  

DEFF Research Database (Denmark)

Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO2 from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame temperature. The flue gas produced thus consists primarily of carbon dioxide and water. Much research on the different aspects of an oxy-fuel power plant has been performed during the last decade. Focus has mainly been on retrofits of existing pulverized-coal-fired power plant units. Green-field plants which provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy-fuel process and focuses particularly on the combustion fundamentals, i.e. flame temperatures and heat transfer, ignition and burnout, emissions, and fly ash characteristics. Knowledge is currently available regarding both an entire oxy-fuel power plant and the combustion fundamentals. However, several questions remain unanswered and more research and pilot plant testing of heat transfer profiles, emission levels, the optimum oxygen excess and inlet oxygen concentration levels, high and low-temperature fire-side corrosion, ash quality, plant operability, and models to predict NOx and SO3 formation is required.

Toftegaard, Maja BØg; Brix, Jacob

2010-01-01

242

Fuel cell technology  

International Nuclear Information System (INIS)

Fuel cell technology is receiving significant attention in recent years because of its potential application as a highly efficient electric power generation system with low environmental impact. There are four fuel cell categories that are currently in use or under development: polymer electrolyte membrane fuel cells (SPFC), phosphoric acid fuel cells (PAFC), molten carbonate fuel cells (MCFC) and solid oxide fuel cells (SOFC). The phosphoric acid is the only available technology that has been produced in approaching series production and has already accumulated a significant in-service experience. It is generally accepted that fuel cells are a viable alternative to the use of internal combustion engine and gas turbines for the electric power generation in the range of capacity from a few kW up to tens MW. At high power ranges (above 20 MW) fuel cells have to compete with well-established technologies, such as gas turbines, that have made great gains over the last decade, in term of efficiency, emission performance and capital cost. Presently the cost of fuel cell plants are still high, but the major technology development companies indicate that the prices will be driven down to the required level (1500 $/kW) by 2000, both through technology refinements and increase of production volume. Market analysis indicates that in Italy the fuel cells could find early applications in three primary areas: dispersed-type power by electric utilities, small-scale cogeneration in rc utilities, small-scale cogeneration in residential and industrial applications and electric transportation

243

Nuclear fuel storage  

International Nuclear Information System (INIS)

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

244

Nuclear fuel rods  

International Nuclear Information System (INIS)

Purpose : To improve the reliability of fuels by reducing the possibility of stress corrosion cracks in a fuel can and decreasing the fuel element failures in a case where stresses are exerted on the fuel can due to the fuel-cladding interaction in corrosive gases. Constitution : The frequency of generating the brittlement crackings in a fuel can is reduced by adjusting the direction of zirconium single crystals constituting the fuel can so that the circumferential direction of the fuel can, that is, the maximum stress exerting direction and the C-axis direction of the zirconium single crystals may be made different from each other as much as possible. Specifically, a hot-extruded material pipe of a zirconium alloy is rolled while applying twisting fabrication forcibly in a certain direction at the intermediate rolling step and the final rolling step. The fuel can thus obtained has a large resistance to the stress corrosion cracks and permits a greater deformation. That is, the fuel can has a great resistance to the brittlement crackings. (Yoshihara, H.)

245

Fuel safety research 1999  

Energy Technology Data Exchange (ETDEWEB)

In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-07-01

246

Method of reactor fueling  

International Nuclear Information System (INIS)

Purpose: To improve the fuel burnup degree and decrease the amount of radioactive wastes in a newly installed nuclear reactors. Method: The reactor core at the initial stage in a newly installed nuclear reactor comprises fuel assemblies and control rods, in which the fuel assemblies are usually supported by four in one set on fuel support metals. Spent fuel assemblies from other existent nuclear reactors and new fuel assemblies with the uranium-235 enrichment degree of about 2.43 % are used as the fuel assemblies constituting the reactor core. The spent fuel assemblies are arranged at the outermost circumferential range facing to the reflector and the central region of the reactor core. The spent fuel assemblies in the central region are arranged with the rotational symmetry in the reactor core and disposed such that they are in adjacent each by one with all of the control rods disposed in the central region. Further, the new fuel assemblies are arranged only at the central region. (Kawakami, Y.)

247

Nuclear fuel activities in Belgium  

International Nuclear Information System (INIS)

In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes

248

Spent fuel transporting cask  

International Nuclear Information System (INIS)

Purpose: To lower the degree of exposure of workers to radiation during the transport of spent fuel and also to improve transport efficiency by increasing the spent fuel holding capacity of the container. Method: A deplated uranium metal is interposed as a gamma ray shielding body between the inner and outer tubes of the shell of the spent fuel transport cask for the purpose of lessening the degree of exposure of workers to radiation by utilizing the excellent gamma ray shielding performance of the depleted uranium metal. Furthermore, the wall thickness of the whole container can be made thinner than casks in conventional use. Therefore, the cask capacity for holding the spent fuel can be increased and the transport efficiency improved. In addition, a large volume of depleted uranium metal is produced in the process of manufacturing fuel to be used for atomic power generation and in the process of re-processing the spent fuel, and can be effectively utilized. (Takahashi, M.)

249

HTR fuel manufacturing experience  

International Nuclear Information System (INIS)

The development of the HTR line promises the availability of a number of technologies which can be used in many area of energy supply. Special properties of nuclear energy exploitation in high-temperature reactors include economical uranium consumption and lower pollution of the environment. Fuel cycle, design and irradiation performance requirements impose restraints on the fuel elements fabrication processes. Both kernel and coating fabrication processes are flexible enough to adapt to the needs of the various existing and proposed high temperature gas-cooled reactors. Extensive experience has demonstrated that fuel kernels with excellent sphericity and uniformity can be produced by wet chemical processes. Similarly experience has shown that the various multilayer coatings can be produced to fully meet design and specification requirements. In a comprehensive qualification program for fuel elements the low failure fraction of coated fuel particles, optimal matrix behavior and the required fission product retention of integral fuel elements was successfully demonstrated

250

Nuclear fuel element  

International Nuclear Information System (INIS)

Object: To provide an arrangement wherein when a fuel pellet within a clad tube is applied with a stress or strain, such can be allowed and failure due to deformation can be prevented. Structure: In a structure wherein a cylindrical fuel pellet including oxide of nuclear fuel material is charged into a sealed clad tube, hardness in an outer peripheral portion of the fuel pellet is made lower than hardness in the center thereof and oxygen concentration in an outer peripheral portion of the fuel pellet is made higher than that of the center thereof. A region, of which rate of bond between the oxygen and the nuclear fuel material is more than 2.05, is present over 15% from the outer periphery of pellet radius. (Yoshihara, H.)

251

Nuclear fuel assembly  

International Nuclear Information System (INIS)

Purpose: To minimize the gap between a wrapper tube and a fuel rod, as well as reduce the swelling stress exerted on the fuel rod. Constitution: The portion of the wrapper tube corresponding to the highest temperature portion in the fuel rod is made of a substance of greater swelling than other portions. For example, in the wrapper tube made of stainless steel through 20% cold working, the degree of cold working for the highest temperature portion is reduced to about 10%. Since greater swelling is resulted in the portion with lower cold working degree, if a fuel rod expands during reaction operation, the portion of the wrapper tube corresponding thereto swells remarkably, so that the gap between the wrapper tube and the fuel rod is not decreased. This allows to reduce the gap between the wrapper tube and the fuel rod enabling to decrease the vibration caused by the flow of the coolant. (Ikeda, J.)

252

Nuclear fuel assemblies  

International Nuclear Information System (INIS)

Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

253

Nuclear fuel cycle  

International Nuclear Information System (INIS)

The elaboration of nuclear fuels starting with raw materials requires a series of complex transformations. The spent fuels also contains some energetic materials, sometimes reusable, and radioactive residual products and must be processed using appropriate treatments before definitive disposal. The nuclear fuel cycle is this succession of processing operations from the uranium mine to the reactor and from the reactor to the radioactive waste ultimate storage facility. This paper briefly describes the processes involved at each step of the fuel cycle and gives a general idea of the industrial installations devoted to uranium production and fuel cycle services worldwide. The situation of the fuel cycle international market and its economical aspects are also briefly analyzed. (J.S.)

254

Nuclear fuel rods  

International Nuclear Information System (INIS)

Purpose: To avoid reduction of fracture stresses in fuel cans by applying a beryllium-copper thin layer to the inner surface of the fuel cans. Constitution: A beryllium-copper thin layer is plated and sintered onto the oxide film on the inner circumferential surface of a cylindrical fuel can made of zirconium alloy and formed with the oxide film at its inner and outer circumferential surfaces by way of autoclaving. The fuel can thus formed with the beryllium-copper thin layer on its inner circumferential surface is welded, at its lower opening, with an end plug, charged with nuclear fuel pellets, replaced with helium, mounted with a pressor plate, a plenum spring or the like and finally welded with an upper end plug to constitute a fuel rod. (Kawakami, Y.)

255

Fuel emissivity (FEMISS)  

International Nuclear Information System (INIS)

The report describes work which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. It is an interim addition to work previously published in the Materials Properties (MATPRO) Handbook and will replace Section A-3 (Fuel Emissivity) of the MATPRO-11 handbook. This update of the fuel emissivity subcode includes new data and an estimate of the standard error to be expected with the subcode. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) analytical programs such as the Fuel Rod Analysis Program--Steady-State (FRAPCON-1) code and the Fuel Rod Analysis Program--Transient (FRAP-T4) code. This work is being performed as part of a broad effort to develop and verify analytical models capable of describing nuclear fuel rod behavior

256

Nuclear fuel cycle standards  

International Nuclear Information System (INIS)

The nuclear fuel cycle standard provides the framework, by using a matrix, for the identification and classification of areas where consensus standards potentially will be needed. The standard provides guidance for establishing priorities, making unambiguous assignments, and reporting and coordinating progress. The standard arranges the nuclear fuel cycle complex into manageable packages so that standards work can be developed by a number of groups simultaneously, or in a disparate time frame, in response to the urgency of the need. The master matrix is defined as the array of operations included in the fuel cycle for nuclear reactors. These operations start with the mining, milling, and fuel fabrication; continue with the transportation of spent fuel after discharge from a reactor; and culminate with the shipment of waste to a repository, of refabricated fuel to a reactor station, and of recycled uranium to an enrichment plant

257

Spiral cooled fuel nozzle  

Science.gov (United States)

A fuel nozzle for delivery of fuel to a gas turbine engine. The fuel nozzle includes an outer nozzle wall and a center body located centrally within the nozzle wall. A gap is defined between an inner wall surface of the nozzle wall and an outer body surface of the center body for providing fuel flow in a longitudinal direction from an inlet end to an outlet end of the fuel nozzle. A turbulating feature is defined on at least one of the central body and the inner wall for causing at least a portion of the fuel flow in the gap to flow transverse to the longitudinal direction. The gap is effective to provide a substantially uniform temperature distribution along the nozzle wall in the circumferential direction.

Fox, Timothy; Schilp, Reinhard

2012-09-25

258

Fuel consolidation demonstration program  

International Nuclear Information System (INIS)

EPRI, Northeast Utilities, Baltimore Gas and Electric, the US Department of Energy and Combustion Engineering are engaged in a program to develop a system for consolidating spent fuel and a method of storing the consolidated fuel in the spent fuel storage pool which is licensable by the US Nuclear Regulatory Commission. Fuel consolidation offers a means of substantially increasing the capacity of spent fuel storage pools. This is a final report of the Fuel Consolidation Demonstration Program. It provides a review of the overall program, a summary of the results obtained, the lessons learned, and an assessment of the present status of the consolidation system developed in the program. 7 refs., 15 figs., 5 tabs

259

Nuclear fuel element  

International Nuclear Information System (INIS)

This invention concerns nuclear fuel elements suitable to the increase of the burnup degree of nuclear fuels. Fuel pellets each having metal rod protrusions on both end faces and metal thin films formed at the outer surface of the pellet, and metal disks each having a central aperture for inserting the metal rod protrusion are alternately combined and stacked and loaded in a metal fuel can. The metal rods and the metal thin films for the fuel pellet are made of tungsten, molybdenum, tantalum or alloys thereof. Heat generated from the metal rod is conducted by way of the metal thin film and the metal disk to the outer circumferential portion and rapidly dissipated. As a result, it is possible to prevent temperature elevation at the central portion of fuel pellets, suppress the release of FP gases, reduce the swelling and the pellet-cladding tube mechanical interactions. (I.N.)

260

Nuclear fuel element  

International Nuclear Information System (INIS)

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

261

Nuclear fuel element  

International Nuclear Information System (INIS)

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

262

The Tarapur fuel controversy  

International Nuclear Information System (INIS)

As a consequence of India's peaceful nuclear explosion at Pokhran in 1974, the United States started delaying the shipments of enriched uranium fuel supplies in order to pressurise India to sign the NPT treaty which India has refused to do so on account of its discriminatory nature. According to the 1963 agreement, the USA is supposed to supply the enriched uranium fuel for Tarapur Atomic Power Station (TAPS) for its entire life time i.e. up to 1993. With the uncertainty in fuel supplies, the Tarapur reactor units are being run at a low capacity factor to conserve fuel. In case the USA abrogates the treaty, India intends to run TAPS on mixed oxide fuel part of which will come from reprocessing of spent fuel of TAPS. Cost of power production will go up. Other problems, particularly the problem of reactor safety will have to be closely looked into. (M.G.B.)

263

Hydrogen - the new fuel  

International Nuclear Information System (INIS)

The energy resources of the planet are discussed. It is pointed out that Hydrogen is one of the most valuable alternatives of classical fuels. The opinion of most of the political and technical authorities from all over the world on this point is cited. They have discussed the possible applications of Hydrogen as fuel for internal combustion engines as well as chemical fuel in the so called 'fuel cells'. It was pointed out that the use of Hydrogen in fuel cells is more prospective alternative for traction purposes, for reserve sources, etc. The most prospective types of fuel cells are considered at the present moment. The methods of Hydrogen production and infrastructure of functioning Hydrogen energetics are also discussed. (authors)

264

Nuclear fuel rod  

International Nuclear Information System (INIS)

Fuel pellets having a low transuranium element content and fuel pellets having a high transuranium element content are contained in a cladding tube. The fuel pellets having a high transuranium content are disposed in an upper portion of a fuel rod. The fuel pellets having a high transuranium content comprises two regions of different transuranium element contents, in which the transuranium element content is increased at a central region than the peripheral region in adjacent with the cladding tube. In this case, the transuranium element is constituted with one or more of neptunium-237, americium-241, and americium-243. The transuranium elements are annihilated effectively by using the fact that the neutron spectrum are harder in an upper portion of the reactor core and also in a central region of the fuel pellet. (I.N.)

265

Fuel Cells and Biogas  

OpenAIRE

This thesis concerns biogas-operated fuel cells. Fuel cell technology may contribute to more efficient energy use, reduce emissions and also perhaps revolutionize current energy systems. The technology is, however, still immature and has not yet been implemented as dominant in any application or niche market. Research and development is currently being carried out to investigate whether fuel cells can live up to their full potential and to further advance the technology. The research of thesi...

Hedstro?m, Lars

2010-01-01

266

Fuel assembly spacer grid  

International Nuclear Information System (INIS)

The patent concerns grids for sub-assemblies of nuclear reactors. The grid comprises a plurality of cells, with each cell receiving one fuel pin. The fuel pin is engaged by two inwardly directed projections within the cell, the projections in successive grids along the pins being displaced by 1200 in the same sense. Such an arrangement facilitates dismantling of the irradiated sub-assembly prior to reprocessing of the fuel. (U.K.)

267

Fuel safety research 2001  

International Nuclear Information System (INIS)

The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

268

Spent fuel storage  

International Nuclear Information System (INIS)

To begin with, the author explains the reasons for intermediate storage of fuel elements in nuclear power stations and in a reprocessing plant and gives the temperature and radioactivity curves of LWR fuel elements after removal from the reactor. This is followed by a description of the facilities for fuel element storage in a reprocessing plant and of their functions. Futher topics are criticality and activity control, the problem of cooling time and safety systems. (HR)

269

Liquid fuel cells  

OpenAIRE

The advantages of liquid fuel cells (LFCs) over conventional hydrogen–oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

Soloveichik, Grigorii L.

2014-01-01

270

Nuclear fuel assembly  

International Nuclear Information System (INIS)

A nuclear fuel assembly comprises a cluster of elongated fuel, retained parallel and at the nodal points of a square network by a bottom supporting plate and by spacing grids. The supporting plate is connected to a top end plate via tie-rods which replace fuel pins at certain of the nodal points of the network. The diameter of the tie-rods is equal to that of the pins and both are slidably received in the grids

271

Plutonium fuel program  

International Nuclear Information System (INIS)

The project is concerned with developing an advanced method to produce nuclear reactor fuels. Since 1968 EIR has worked successfully on the production of uranium-plutonium mixed carbide using wet gelation chemistry. An important part of the development is irradiating the fuel in materials test reactors and evaluating its performance. During 1979 the programme continued with principal activities of fuel fabrication development, preparation for irradiation testing, performance evaluation, and modelling and plant engineering. (Auth.)

272

What fuel for SFRs?  

International Nuclear Information System (INIS)

Fuel for a fourth-generation sodium-cooled fast reactor will have to meet specifications involving novel requirements, in terms of density, and thermic characteristics, and entailing further, new consequences as regards the core, assemblies, and fuel pins. Over the next two years, researchers are set to define fuels that will afford the capability to meet the specifications for reactors of the latest generation in this technology line. (author)

273

Fuel safety research 2001  

Energy Technology Data Exchange (ETDEWEB)

The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-11-01

274

Find a Fuel Game  

Science.gov (United States)

An educational game where students choose vehicles with different propulsion systems and learn about the differences between these systems. To progress in the game, students must capture the fuel type for the propulsion system of the vehicle they chose. Propulsion systems and fuel types include the internal combustion engine (gasoline),  flex-fuel ethanol (E85), hybrid electric (gasoline and electricity), and plug-in hybrid electric (electricty or both gasoline and electricity).

General Motors

275

Transient fuel melting  

International Nuclear Information System (INIS)

The observation of micrographic documents from fuel after a CABRI test leads to postulate a specific mode of transient fuel melting during a rapid nuclear power excursion. When reaching the melt threshold, the bands which are characteristic for the solid state are broken statistically over a macroscopic region. The time of maintaining the fuel at the critical enthalpy level between solid and liquid is too short to lead to a phase separation. A significant life-time (approximately 1 second) of this intermediate ''unsolide'' state would have consequences on the variation of physical properties linked to the phase transition solid/liquid: viscosity, specific volume and (for the irradiated fuel) fission gas release

276

FAILED FUEL DISPOSITION STUDY  

International Nuclear Information System (INIS)

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and ns from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This alternative does not afford the ability to inspect the damaged fuel prior to placing it into storage. This alternative would require a much more extensive analyses to revise the 200 Area ISA FSAR for this fuel pin condition and storage configuration crediting the ID-69 container for retrievability and the core component container (CCC) as the primary confinement boundary in addition to the canning function

277

The nuclear fuel cycle  

International Nuclear Information System (INIS)

After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

278

Fuel cells - a perspective  

International Nuclear Information System (INIS)

Unfortunately, fuel cell publicity conveys expectations and hopes that are often based on uncritical interpretations of the underlying science. The aim here is to use that science to analyse how the technology has developed and what can realistically be delivered by fuel cells. There have been great achievements in fuel cell technology over the past decade, with most types reaching an advanced stage of engineering development. But there has been some muddled thinking about one critical aspect, fuel cell energy efficiency. The 'Carnot cycle' argument, that fuel cells must be much more efficient than heat engines, is a red herring, of no help in predicting real efficiencies. In practice, fuel cells are not always particularly efficient and there are good scientific reasons for this. Cost reduction is a big issue for fuel cells. They are not in principle especially simple devices. Better engineering and mass production will presumably bring costs down, but because of their inherent complexity there is no reason to expect them to be cheap. It is fair to conclude that predictions of fuel cells as commonplace components of energy systems (including a hydrogen economy) need to be treated with caution, at least until major improvements eventuate. However, one type, the direct methanol fuel cell, is aimed at a clear existing market in consumer electronics

279

TRIGA fuel element instrumentation  

International Nuclear Information System (INIS)

To upgrade the reliability of TRIGA fuel temperature measurements, the instrumented element has been redesigned to increase thermocouple life. The primary requirement of the new design is to improve thermocouple reliability while maintaining or exceeding the time response of the existing system. The thermocouples are located in the fuel at three elevations (radially identical) spanning the fuel center. Additional instrumentation, in the form of a self-powered neutron flux monitor is internally applied along the fuel center line. The Atomics Internation thermocouple design and implementation are described

280

HTGR fuel performance basis  

International Nuclear Information System (INIS)

The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 16000C, and complete fuel failure occurs at 26600C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

281

A perfect fuel supplier  

International Nuclear Information System (INIS)

WWER fuel market is dominated by the Russian fuel vendor JSC TVEL. There have been attempts to open up the market also for other suppliers, such as BNFL/Westinghouse for Finland, Czech Republic, and Ukraine. However, at the moment it seems that JSC TVEL is the only real alternative to supply fuel to WWER reactors. All existing fuel suppliers have certified quality management systems which put a special emphasis on the customer satisfaction. This paper attempts to define from the customer's point of view, what are the important issues concerning the customer satisfaction. (author)

282

Fuel cell stack arrangements  

Science.gov (United States)

Arrangements of stacks of fuel cells and ducts, for fuel cells operating with separate fuel, oxidant and coolant streams. An even number of stacks are arranged generally end-to-end in a loop. Ducts located at the juncture of consecutive stacks of the loop feed oxidant or fuel to or from the two consecutive stacks, each individual duct communicating with two stacks. A coolant fluid flows from outside the loop, into and through cooling channels of the stack, and is discharged into an enclosure duct formed within the loop by the stacks and seals at the junctures at the stacks.

Kothmann, Richard E. (Churchill Boro, PA); Somers, Edward V. (Murrysville, PA)

1982-01-01

283

Nuclear fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To eliminate the possibility of fuel raptures even if a great heat elongation is resulted to a fuel rod, by providing an external spring having a large compression amount capable of coping with a large heat elongation of the fuel rod. Constitution: An external spring fitted to the outer circumference of a support shaft above the upper end plug and between the flange at the bottom of the upper end plug and an upper tie plate is made of a conical compressive spring being capable of compressing to a height equal to the diameter of the material. Accordingly, when a fuel assembly burns in a reactor and a fuel rod is thermally extended, the outer spring as the conical compressive coil spring shows a greater compressive strength than the usual external spring and the spring can be shortened to the height equal to the material diameter if the fuel rod causes a large thermal elongation. Consequently, since the fuel rod can freely be extended not exposed to the direct and great elongation force of the fuel rod but only under the extending tendency of the external spring, whereby the damage for the fuel rod can be prevented. (Yoshihara, H.)

284

Fuel Cell Demonstration Program  

Energy Technology Data Exchange (ETDEWEB)

In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance, installation, and decommissioning the total project budget was approximately $3.7 million.

Gerald Brun

2006-09-15

285

Fuel assembly reconstitution  

International Nuclear Information System (INIS)

Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

286

Cracked fuel mechanics  

International Nuclear Information System (INIS)

Fuel pellets undergo thermally induced cracking during normal reactor operation. Some fuel performance codes have included models that address the effects of fuel cracking on the fuel rod thermal and mechanical behavior. However, models that rely too heavily on continuum mechanics formulations (annular gaps and solid cylindrical pellets) characteristically do not adequately predict cladding axial elongations. This paper presents a modelling concept and a set of measurable parameters that have been shown to improve the prediction of the mechanical behavior of cracked fuel-cladding systems without added computational expense. The transition from classical annular gap/cylindrical pellet models to modified bulk properties and further to local behavior for cracked fuel systems is discussed. The results of laboratory experiments to verify these modeling parameters are shown. Data are also presented from laboratory experiments on unirradiated and irradiated rods which show that fuel rod mechanical response depends on fuel fragment size. The impact of these data on cracked fuel behavior and failure modeling is also discussed. (author)

287

Nuclear fuel assembly  

International Nuclear Information System (INIS)

Purpose: To increase the maximum allowable thermal power of nuclear fuel assemblies than usual in BWR type reactors. Constitution: A plurality of spacers are disposed to a fuel rod at a predetermined interval in the direction of the fuel rods, in which the flow of the coolants is resisted to worsen the cooling effect, as well as the thermal power exceeds the cooling performance near the upstream of the spacers to rise the temperature of the cladding tube till the burning damage may possibly be resulted. In view of the above, hollow fuel pellets are charged to a portion of the fuel rod from 98.5 to 100.5 % length assuming the distance from the lower surface of the fuel pellet charged at the lowermost end of the fuel rod to the upper end of each of the spacers as 100 %. In this way, since the hollow fuel pellets with a lower thermal power are disposed near the upper side of the spacers, the burning accident to the cladding tube at this portion can be avoided. Accordingly, the maximum allowable thermal power of the nuclear fuel assembly can be increased. (Kawakami, Y.)

288

ITER fuel cycle  

International Nuclear Information System (INIS)

Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

289

Nuclear fuel cycle options  

International Nuclear Information System (INIS)

Presently, industrial maturity can be claimed for two fuel cycle strategies, viz. the 'Once Through Fuel Cycle' (OTC), and the 'Reprocessing Fuel Cycle' (RFC) in which plutonium and very limited uranium quantities are being recycled. It is helpful to recall some key data that set the stage for any discussion of fuel cycle options: 1. Worldwide, the annual spent fuel discharge is in the range of 10500-11000 t heavy-metal (HM), while the industrial reprocessing capacity amounts to ? 5000 t HM (OECD NUCLEAR ENERGY AGENCY, Accelerator-driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles: a Comparative Study, Paris, 2002). Hence, less than 1/2 of the discharged spent fuel can be processed. 2. Worldwide, the cumulative inventory of stored spent fuel is estimated to be ? 190000 t HM, and the amount of reprocessed spent fuel is estimated to be ? 70000 t HM. The latter inventory has been transformed into high-level waste (HLW) and spent light water reactor (LWR) mixed uranium-plutonium oxide (MOX) fuel. Considering the relatively low uranium ore prices, this situation is expected to continue over the next few decades. 3. Therefore, it is likely that the need for repository space will increase accordingly. Taking the Yucca Mountain project (63000 t HM capacity) as reference repository, the present worldwide inventories would require three repositories for the spent fuel, and one for the HLW. For the USA alone (OTC strategy), assuming a life time ext strategy), assuming a life time extension of the present nuclear reactors to 60 years, and no new reactors, the capacity of Yucca Mountain will be exceeded by ? 2050. Considering aforementioned data and given the strong public opposition to the construction of geologic repositories, it is understandable that over the last decade or so, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies aiming at reducing the amount of long-lived radioactive waste through the introduction of innovative/advanced fuel cycles, including transmutation in fission reactors or dedicated reactors like Accelerator Driven Systems (ADS) of Fusion/Fission systems. After briefly reviewing the established fuel cycle options (OTC and RFC), the paper will first summarize the main features of innovative fuel cycle options (including recycle/partitioning aspects, fuel fabrication, spent fuel management, and reactor design), and then give a broad overview of national and international research and technology development activities, including IAEA's own, in this area. (author)

290

Nuclear fuel assembly  

Energy Technology Data Exchange (ETDEWEB)

The present invention provides a fuel assembly for a BWR type reactor capable of improving the fuel economy by improving a neutron utilizing efficiency, improving the reactor core stability by the reduction of pressure loss in the fuel assembly and improving the thermal margin relative to boiling transfer. The fuel assembly comprises two kinds of fuel rods having different lengths and water rod disposed in lattice like manner. In this case, the water rods are disposed at a central portion of the transverse cross section of the assembly. The total cross sectional area of the water rods is that of four or more fuel rods. Short fuel rods, among fuel rods, are disposed at corners of the outermost circumference of the cross section or the corner of the second layer from the outer side of the assembly and not being in adjacent with the water rods. With such a constitution, the water rods compensate portions with insufficient moderators at central portion of the cross section of the assembly to moderate neutrons thereby enhancing the utilizing efficiency. Since the short fuel rods are disposed at the corners of the outermost circumference, elevation of reactivity upon cold temperature and the reactivity upon void change can be suppressed. (I.S.)

Bessho, Yasunori; Aoyama, Motoo; Ishii, Yoshihiko; Uchikawa, Sadao; Morimoto, Yuichi

1997-08-19

291

Nuclear fuel assembly  

International Nuclear Information System (INIS)

Purpose: To flatten the axial power distribution by reducing the reactivity change accompanying the change of the moderator density in a nuclear fuel assembly in which fuel rods comprising 235U as the nuclear fuel material are disposed densely. Constitution: In a nuclear fuel assembly charged to a BWR type reactor, a fuel rod is axially divided into a plurality of regions, in which total loading amount of 235U and 238U are made smaller in the upper portion and made greater in the lower portion, and the enrichment degree of 235U is made higher in the upper portion and lower in the lower portion, so that the ratio of the area occupied by the coolants and that by the fuels is made less than 1.5. Further, a plurality of spacers are disposed in the axial direction of a bundle of fuel rods to prevent contact of adjacent fuel rods from each other and ensure the channels between each of the fuel rods for flowing coolants. This can flatten the axial power distribution even if there is any density difference in the moderaters. (Takahashi, M.)

292

Protocol Fuel Mix reporting  

International Nuclear Information System (INIS)

The protocol in this document describes a method for an Electricity Distribution Company (EDC) to account for the fuel mix of electricity that it delivers to its customers, based on the best available information. Own production, purchase and sale of electricity, and certificates trading are taken into account. In chapter 2 the actual protocol is outlined. In the appendixes additional (supporting) information is given: (A) Dutch Standard Fuel Mix, 2000; (B) Calculation of the Dutch Standard fuel mix; (C) Procedures to estimate and benchmark the fuel mix; (D) Quality management; (E) External verification; (F) Recommendation for further development of the protocol; (G) Reporting examples

293

Breeder fuel reprocessing  

International Nuclear Information System (INIS)

Two definite specific subjects on the mechanical operations of reprocessing are described: 1) decladding of the fuels, as used since 1973 at Marcoule with the Phenix reactor, by saving and stripping the chips on the cladding in order to recover the bundle of fuel rods, 2) opening the transport cases of Super Phenix fuel in sodium, with a CO2 power laser, in a hot cell work configuration, namely on irradiated materials, in a shielded room with remote controlled work as envisaged for the so called reference solution (transport of fuels in liquid sodium filled cases)

294

Fuel channel performance  

International Nuclear Information System (INIS)

This paper summarizes the performance of fuel channels in CANDU reactors. The evolution of the overall fuel channel design and the modifications to individual components are described. The main fuel channel component, the pressure tube, is subject from service conditions, to changes in three principal factors, dimensions, properties and composition, each of which can affect performance or life of the tube. The changes that occur are reviewed briefly. The performance of the channels from the view point of operating problems and replacement experience show the relatively low man-rem expenditure associated with fuel channel replacement. The report concludes with an outline of channel design development

295

Nuclear fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To substantially reduce the effective multiplication factor of neutrons thereby increase the reactor shutdown margin in nuclear fuel assemblies of BWR type reactors. Constitution: The nuclear fuel assembly comprises an upper connection plate and a lower connection plate for supporting upper and lower ends of fuel rods respectively, a plurality of spacers disposed at the midway of the fuel rods for regulating the gaps between the fuel rods and a channel box covering them. The fuel rods are charged with uranium dioxide and gadolinia. A portion of a plurality of fuel rods is mixed with gadolinia with the distribution of the concentration being varied in the axial direction. For the distribution of the gadolinia concentration, the boundary between the upper region and the central region situates within a range to 16/24 ? 22/24 from the lowermost end of the effective length of fuels, while the boundary between the central region and the lower region situates within a range to 8/24 ? 14/24 from the lowest end of the fuel effective length. The uranium enrichment degree can thus be increased to increase the reactor operation period and the reactor shutdown margin. (Horiuchi, T.)

296

Nuclear fuel quality assurance  

International Nuclear Information System (INIS)

Full text: Quality assurance is used extensively in the design, construction and operation of nuclear power plants. This methodology is applied to all activities affecting the quality of a nuclear power plant in order to obtain confidence that an item or a facility will perform satisfactorily in service. Although the achievement of quality is the responsibility of all parties participating in a nuclear power project, establishment and implementation of the quality assurance programme for the whole plant is a main responsibility of the plant owner. For the plant owner, the main concern is to achieve control over the quality of purchased products or services through contractual arrangements with the vendors. In the case of purchase of nuclear fuel, the application of quality assurance might be faced with several difficulties because of the lack of standardization in nuclear fuel and the proprietary information of the fuel manufacturers on fuel design specifications and fuel manufacturing procedures. The problems of quality assurance for purchase of nuclear fuel were discussed in detail during the seminar. Due to the lack of generally acceptable standards, the successful application of the quality assurance concept to the procurement of fuel depends on how much information can be provided by the fuel manufacturer to the utility which is purchasing fuel, and in what form and how early this information can be provided. The extent of information transfer is basically set ouf information transfer is basically set out in the individual vendor-utility contracts, with some indirect influence from the requirements of regulatory bodies. Any conflict that exists appears to come from utilities which desire more extensive control over the product they are buying. There is a reluctance on the part of vendors to permit close insight of the purchasers into their design and manufacturing procedures, but there nevertheless seems to be an increasing trend towards release of more information to the purchasers. It appears that the full application of the quality assurance concept in the purchase of fuel and fuel manufacturing services will depend to a large extent on the availability of fuel specification data. On the part of fuel purchasers, there is an obvious interest in getting as many details of fuel specification as possible in order to be able to establish a proper level of control over the quality of their purchases. On the other hand, if such specifications are set up in advance by the purchasers, there are often complaints by the manufacturers that the specifications were set up without proper regard for the latest technical information on fuel performance and for the realities of manufacturing processes and technical capabilities. This problem may be resolved when fuel design activities are properly meshed with a full quality assurance system. Discussions during the seminar showed that the operation of acceptable quality assurance systems is a well-established practice at most of the fuel manufacturers. The fuel purchaser may monitor such a system through quality assurance programme auditing as agreed to the individual vendor-purchaser contracts. In this way confidence may be obtained in the quality of the purchased product. However, it is considered that the further improvement of the relations between fuel manufacturers and purchasers could be achieved through the following actions undertaken at the international level: (1) standardization of fuel specifications and testing procedures; (2) dissemination of information on fuel specifications and their connections with observed fuel failure rate; (3) Establishment of a standardized quality assurance programme for fuel fabrication; (4) establishment of a central information service to assist utility groups in preparing documents and procedures to be used in quality assurance activities

297

Nuclear fuel cycle  

International Nuclear Information System (INIS)

In a nuclear reactor, fuel is where the fission process of heavy uranium or plutonium atoms takes place. This fission process is the source of the heat needed to produce electricity (via a turbine), or the source of energy for other applications. The heavy nuclei must follow a process including several industrial steps before they can be used in the reactor: - extraction of uranium ore, - concentration of the ore and its conversion into gaseous uranium hexafluoride for enrichment, - isotopic enrichment of uranium to increase the proportion of the fissile nuclei (U-235), too low in the natural state, and re-conversion of this fuel into uranium oxide powder (UO2), - fabrication of fuel in the form of 'pellets': small cylinders, approximately 1 cm in length, weighing 7 grams. The pellets are inserted into long metal tubes measuring 4 m in length. These tubes are known as the cladding. The ends are sealed to make fuel rods, bundled into assemblies which are then placed in the reactor core. Approximately 40,000 rods are prepared per nuclear power plant and bundled into 'assemblies' of 264 rods each. It takes 157 fuel assemblies containing a total of 11 million pellets to load a 900 MW nuclear reactor. After this 'front end' phase is completed, the fuel can be used for energy production inside the reactor. Neutrons split the fuel's U-235 nuclei; as this fission takes place, energy is released along with more neutrons, which go on to split other U-235 nuclei, thus creating a chain reaction. The fuel remains approximately four years in the reactor. Afterwards the 'back end' of the fuel cycle begins, including: - temporary storage of spent fuel under water for cooling purposes, - spent fuel management is the last step in the cycle. It is different for an 'open' cycle than a closed cycle. - chemical processing of spent fuel to separate fissile and reusable material, - recycling of reactor-grade plutonium into MOX fuel (mixed oxide fuel), - 'final' waste packaging and vitrification of the most radioactive waste, then waste storage. The closed fuel cycle has been France's industrial choice since the 1980's with the opening of the Areva plant at La Hague and the Melox plant. This example has been followed by Germany, Japan and Switzerland. The United States and China are also seriously considering this option. The CEA has been exploring this research area for nearly half a century, from the first separative chemistry studies in the 1950's to the advanced processes developed today at Marcoule. (authors)

298

TRIGA low enrichment fuel  

International Nuclear Information System (INIS)

Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

299

TRIGA low enrichment fuel  

International Nuclear Information System (INIS)

Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

300

Fuel safety research 2000  

Energy Technology Data Exchange (ETDEWEB)

In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

301

Plasma sprayed and electrospark deposited zirconium metal diffusion barrier coatings  

Energy Technology Data Exchange (ETDEWEB)

Zirconium metal coatings applied by plasma spraying and electrospark deposition (ESD) have been investigated for use as diffusion barrier coatings on low enrichment uranium fuel for research nuclear reactors. The coatings have been applied to both stainless steel as a surrogate and to simulated nuclear fuel uranium-molybdenum alloy substrates. Deposition parameter development accompanied by coating characterization has been performed. The structure of the plasma sprayed coating was shown to vary with transferred arc current during deposition. The structure of ESD coatings was shown to vary with the capacitance of the deposition equipment.

Hollis, Kendall J [Los Alamos National Laboratory; Pena, Maria I [Los Alamos National Laboratory

2010-01-01

302

Failed fuel detector  

International Nuclear Information System (INIS)

A failed fuel detection apparatus is described for a nuclear reactor having a liquid cooled core comprising a gas collection hood adapted to engage the top of the suspect assembly and means for delivering a stripping gas to the vicinity of the bottom of the suspect fuel assembly. (U.S.)

303

Renewable Fuels Association  

Science.gov (United States)

The national trade association for the U.S. fuel ethanol industry, the Renewable Fuels Association (RFA) serves as a vital link between the ethanol industry and the federal government, to promote increased production and use of ethanol through supportive policies, regulations, and research & development initiatives.

304

Nuclear fuel recycling system  

International Nuclear Information System (INIS)

A process is provided for recycling sintered uranium dioxide fuel pellets rejected during fuel manufacture and the swarf from pellet grinding. The scrap material is prepared mechanically by crushing and milling as a high solids content slurry, using scrap sintered UO2 pellets as the grinding medium under an inert atmosophere

305

Nuclear fuel element  

International Nuclear Information System (INIS)

Object: To provide sintered nuclear fuel pellets subject to less strain during high temperature sintering and to prevent damage to tubular clad. Structure: The instant nuclear fuel element comprises a plurality of cylindrical sintered nuclear fuel pellets stacked one above another within a tubular clad in the axial direction thereof and each provided with ring-like grooves formed at opposite ends. As the linear output density of fuel progressively increases with the sintering of the nuclear fuel element, with low linear output density the center of the nuclear fuel pellet is hard and has high coefficient of thermal expansion, so that contact pressure between adjacent pellets is concentrated at the center of the end face. With an intermediate linear output density plastic deformation is produced at the center of the nuclear fuel pellet, so that the contact pressure is dispersed from the center toward the edge. With high linear output density the contact pressure is dispersed by the groove, so that the strain radial of the nuclear fuel pellet is held at a minimum

306

Spent nuclear fuel storage  

International Nuclear Information System (INIS)

When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

307

Mixed oxide fuel development  

International Nuclear Information System (INIS)

This paper describes the success of the ongoing mixed-oxide fuel development program in the United States aimed at qualifying an economical fuel system for liquid metal cooled reactors. This development has been the cornerstone of the US program for the past 20 years and has proceeded in a deliberate and highly disciplined fashion with high emphasis on fuel reliability and operational safety as major features of an economical fuel system. The program progresses from feature testing in EBR-II to qualifying full size components in FFTF under fully prototypic conditions to establish a basis for extending allowable lifetimes. The development program started with the one year (300 EFPD) core, which is the FFTF driver fuel, continued with the demonstration of a two year (600 EFPD) core and is presently evaluating a three year (900 EFPD) fuel system. All three of these systems, consistent with other LMR fuel programs around the world, use fuel pellets gas bonded to a cladding tube that is assembled into a bundle and fitted into a wrapper tube or duct for ease of insertion into a core. The materials of construction progressed from austenitic CW 316 SS to lower swelling austenitic D9 to non swelling ferritic/martensitic HT9. 6 figs., 2 tabs

308

Mixed oxide fuel development  

Energy Technology Data Exchange (ETDEWEB)

This paper describes the success of the ongoing mixed-oxide fuel development program in the United States aimed at qualifying an economical fuel system for liquid metal cooled reactors. This development has been the cornerstone of the US program for the past 20 years and has proceeded in a deliberate and highly disciplined fashion with high emphasis on fuel reliability and operational safety as major features of an economical fuel system. The program progresses from feature testing in EBR-II to qualifying full size components in FFTF under fully prototypic conditions to establish a basis for extending allowable lifetimes. The development program started with the one year (300 EFPD) core, which is the FFTF driver fuel, continued with the demonstration of a two year (600 EFPD) core and is presently evaluating a three year (900 EFPD) fuel system. All three of these systems, consistent with other LMR fuel programs around the world, use fuel pellets gas bonded to a cladding tube that is assembled into a bundle and fitted into a wrapper tube or duct for ease of insertion into a core. The materials of construction progressed from austenitic CW 316 SS to lower swelling austenitic D9 to non swelling ferritic/martensitic HT9. 6 figs., 2 tabs.

Leggett, R.D.; Omberg, R.P.

1987-05-08

309

Alternative Fuels in Transportation  

Science.gov (United States)

The realization of dwindling fossil fuel supplies and their adverse environmental impacts has accelerated research and development activities in the domain of renewable energy sources and technologies. Global energy demand is expected to rise during the next few decades, and the majority of today's energy is based on fossil fuels. Alternative…

Kouroussis, Denis; Karimi, Shahram

2006-01-01

310

Nuclear fuel element  

International Nuclear Information System (INIS)

A lower end plug is divided into a fuel element portion at an upper portion and an element stopping portion at a lower portion. There are four recessed grooves open to the outer circumference in the fuel element portion and there is an engaging portion at the center. The element stopping portion has one recessed groove opening to the outer circumference thereof and has an engaging hole at the center. The fuel element portion and the element stopping portion have securing holes in radial direction respectively in communication with the recessed grooves. The engaging portion of the fuel element portion and the engaging hole of the element stopping portion are engaged to each other and a U-shaped connection piece is inserted to recessed groove and the securing hole of both of them, to connect the fuel element portion and the element stopping portion. When curved fuel elements are taken out of the reactor and then recharged, the fuel element portion and the element stopping portion are connected while changing the circumferential position so as to upset the curved direction. This can facilitate the recharging of the fuel elements. Further, progress of fretting corrosion can be prevented. (I.N.)

311

Nuclear fuel column retainer  

International Nuclear Information System (INIS)

A description is given of a barrier member fixed in the end of a fuel column retaining spring to prevent contact between the retaining spring and the adjacent end plug of the fuel element whereby contamination of the weld between the cladding tube and end plug with retaining spring material is avoided. 12 claims, 5 drawing figures

312

Nuclear fuel rod  

International Nuclear Information System (INIS)

Purpose; To obtain a fuel rod which has high safety with less mutual action between a fuel pollet and a cladding tube at low operating temperature of the pellet. Constitution: This fuel rod is formed of a hollow fuel pellet having a central hode without a gap between a cladding tube and the pellet. The diameter D0 of the central hole is so decided as to become D02 = D22 - D12, where the inner diameter of the tube is represented by D2 and the outer diameter of the pellet is represented by D1. In this manner, the center temperature of the fuel pellet becomes lower by approx. 100 C. FP emission rate becomes approx. 7 to 8% and the expansion of the fuel pellet becomes lower by approx. 0.2% due to the lower center temperature of the pellet, thereby obtaining a fuel rod having small mutual action between the fuel pellet and the cladding tube. (Yoshihara, H.)

313

Nuclear fuel manufacture  

International Nuclear Information System (INIS)

The technologies used to manufacture nuclear fuel from uranium ore are outlined, with particular reference to the light water reactor fuel cycle. Capital and operating cost estimates for the processing stages are given, and the relevance to a developing uranium industry in Australia is discussed

314

Nuclear fuel transportation containers  

International Nuclear Information System (INIS)

The invention discloses an inner container for a nuclear fuel transportation flask for irradiated fuel elements comprising a cylindrical shell having a dished end closure with a drainage sump and means for flushing out solid matter by way of the sump prior to removing a cover

315

The nuclear fuel cycle  

International Nuclear Information System (INIS)

An overview of nuclear fuel cycle technology is presented. The process of uranium-plutonium fuel cycle is shown with a flow chart from uranium mining through conversion, enrichment, fuel fabrication, irradiated fuel storage, fuel reprocessing including mixed UO2-PuO2 reprocessing for fast breeder reactors to waste storage. The uranium resources reasonably assured at the cost under $80/kgU, and the uranium annual production in 1978 in principal countries are tabulated. The technical issues are outlined for each item of uranium minerals and mining, uranium milling, uranium purification, natural uranium conversion including conventional uranium refining processes and Allied Chemical UF6 process, uranium enrichment technologies such as gaseous diffusion, gas centrifuge, separation nozzle process, South African process, French Chemex process and three advanced processes developed by USDOE, namely atomic vapor laser isotope separation process (AVLIS), molecular laser isotope separation process (MLIS) and plasma separation process, conversion of UF6 to UO2, fuel fabrication, irradiation in converter reactor, irradiated fuel storage and reprocessing by purex process. The fast breeder fuel cycle constitutes a feature of handling mixed dioxide with burn-up from 60,000 to 100,000 megawatt-days per ton and depleted uranium dioxide. The waste processing and waste storage are outlined for vitrification and Joule-heated ceramic or vitrification and Joule-heated ceramic process. (Nakai, Y.)

316

Nuclear fuel assembly  

International Nuclear Information System (INIS)

Purpose: To improve the fuel economy, increase the thermal margin to burnout and decrease the pressure loss to thereby improve the reactor core stability in fuel assemblies. Constitution: Moderator rods have been disposed at the center of fuel assemblies to effectively moderate neutrons thereby improving the neutron availability. The cross section of the moderator rod occupying the containing space for the fuel rods at the center of the fuel assembly is made square or like other shape and the occupying area is so adapted that a predetermined condition can be satisfied between the outer diameter of the fuel rod and the lattice arrangement pitch and the unit constituent cells of the spacer at the periphery of the fuel rod is partially depleted. Further, protrusions are formed to the surface of the moderator rod so as to keep a distance between them. In this way, the diameter of the moderating rod is increased to sufficiently moderate neutrons thereby improving the fuel economy, decreasing the strain in the steam weight % distribution and increasing the thermal margin. Further, the pressure loss is decreased as well to improve the stability of nuclear hot water. (Kamimura, M.)

317

Fuel Cell Laboratory Exercise  

Science.gov (United States)

This in-class lab exercise gives students the chance to build a zinc- copper fuel cell out of its component parts. The procedure for the lab is provided along with a graphical representation of what the fuel cell should look like. Several student questions are also included. This document may be downloaded in PDF file format.

318

Fuel sorting evaluation  

International Nuclear Information System (INIS)

An evaluation of functions and requirements associated with sorting fuel from the Hanford K Basins is presented to support design issue resolution decisions for achieving interim fuel storage. Potential requirements are recommended for implementation in design activities. The recommendations are provided as input to a management decision process where decisions are finalized and the sorting issue is closed

319

Fuel design and engineering  

International Nuclear Information System (INIS)

The essential aspects of the design and engineering of fuel assemblies for LWR reactors are outlined, and the major criteria to be met by the materials used are given. The fuel rods must be mechanically designed to withstand many stresses which are shortly dealt with here. (RB)

320

PLATINUM AND FUEL CELLS  

Science.gov (United States)

Platinum requirements for fuel cell vehicles (FCVS) have been identified as a concern and possible problem with FCV market penetration. Platinum is a necessary component of the electrodes of fuel cell engines that power the vehicles. The platinum is deposited on porous electrodes...

321

Nuclear reactor fuel element  

International Nuclear Information System (INIS)

The grid-shaped spacer for PWR fuel elements consists of flat, upright metal bars at right angles to the fuel rods. In one corner of a grid mesh it has a spring with two end parts for the fuel rod. The cut-outs for the end parts start from an end edge of the metal bar parallel to the fuel rods. The transverse metal bar is one of four outer metal bars. Both end parts of the spring have an extension parallel to this outer metal arm, which grips a grid mesh adjacent to this grid mesh at the side in one corner of the spacer and forms an end part of a spring for the fuel rod there on the inside of the outer metal bar. (HP)

322

Cornell Fuel Cell Institute  

Science.gov (United States)

The Fuel Cell Institute at Cornell University takes "An Advanced Materials Approach to Fuel Cell Technologies." Materials experts at the Institute are examining ways to improve the efficiency of the main components of a low temperature (< 150 Degrees C) fuel cell and adapt reformer catalysts for low temperature operation. The website reviews some of the basics on fuel cells and identifies the remaining research challenges, including questions regarding the materials used in the main components of a fuel cell, such as the anode, the cathode, membrane assembly and, the reformer. These components and their research approach are described further, along with pictures and diagrams to illustrate the processes. Recent publications are available to download.

323

Fuel cell water transport  

Science.gov (United States)

The moisture content and temperature of hydrogen and oxygen gases is regulated throughout traverse of the gases in a fuel cell incorporating a solid polymer membrane. At least one of the gases traverses a first flow field adjacent the solid polymer membrane, where chemical reactions occur to generate an electrical current. A second flow field is located sequential with the first flow field and incorporates a membrane for effective water transport. A control fluid is then circulated adjacent the second membrane on the face opposite the fuel cell gas wherein moisture is either transported from the control fluid to humidify a fuel gas, e.g., hydrogen, or to the control fluid to prevent excess water buildup in the oxidizer gas, e.g., oxygen. Evaporation of water into the control gas and the control gas temperature act to control the fuel cell gas temperatures throughout the traverse of the fuel cell by the gases.

Vanderborgh, Nicholas E. (Los Alamos, NM); Hedstrom, James C. (Los Alamos, NM)

1990-01-01

324

Methanol commercial aviation fuel  

International Nuclear Information System (INIS)

Southern California's heavy reliance on petroleum-fueled transportation has resulted in significant air pollution problems within the south Coast Air Basin (Basin) which stem directly from this near total dependence on fossil fuels. To deal with this pressing issue, recently enacted state legislation has proposed mandatory introduction of clean alternative fuels into ground transportation fleets operating within this area. The commercial air transportation sector, however, also exerts a significant impact on regional air quality which may exceed emission gains achieved in the ground transportation sector. This paper addresses the potential, through the implementation of methanol as a commercial aviation fuel, to improve regional air quality within the Basin and the need to flight test and demonstrate methanol as an environmentally preferable fuel in aircraft turbine engines

325

Hydrogen Fuel Quality  

Energy Technology Data Exchange (ETDEWEB)

For the past 6 years, open discussions and/or meetings have been held and are still on-going with OEM, Hydrogen Suppliers, other test facilities from the North America Team and International collaborators regarding experimental results, fuel clean-up cost, modeling, and analytical techniques to help determine levels of constituents for the development of an international standard for hydrogen fuel quality (ISO TC197 WG-12). Significant progress has been made. The process for the fuel standard is entering final stages as a result of the technical accomplishments. The objectives are to: (1) Determine the allowable levels of hydrogen fuel contaminants in support of the development of science-based international standards for hydrogen fuel quality (ISO TC197 WG-12); and (2) Validate the ASTM test method for determining low levels of non-hydrogen constituents.

Rockward, Tommy [Los Alamos National Laboratory

2012-07-16

326

Vibration of fuel bundles  

International Nuclear Information System (INIS)

Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code, AMASS, is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension, and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods

327

ALMR fuel cycle economics  

International Nuclear Information System (INIS)

The U.S. Advanced Liquid Metal Reactor concept, ALMR, is expected to use recycled metal fuel composed of plutonium, uranium and zirconium. Projections of fuel cycle costs for the ALMR indicate that they will be from 5 to 6 mills/kWh (constant 1987 dollars), which is less than or equal to current light water reactor fuel costs. Since the ALMR fuel costs are not sensitive to uranium ore prices, these costs will be more stable than those of future light water reactors. The ALMR also has the potential to fission actinides from both light water and liquid metal reactor fuels. This ability can reduce the time that high-level radioactive waste are hazardous from millions to hundreds of years with about a 1 mill/kWh increase in cost. (orig.)

328

Fuel cells : emerging markets  

International Nuclear Information System (INIS)

This presentation highlighted the findings of the 2009 review of the fuel cell industry and emerging markets as they appeared in Fuel Cell Today (FCT), a benchmark document on global fuel cell activity. Since 2008, the industry has seen a 50 per cent increase in fuel cell systems shipped, from 12,000 units to 18,000 units. Applications have increased for backup power for datacentres, telecoms and light duty vehicles. The 2009 review focused on emerging markets which include non-traditional regions that may experience considerable diffusion of fuel cells within the next 5 year forecast period. The 2009 review included an analysis on the United Arab Emirates, Mexico, Brazil and India and reviewed primary drivers, likely applications for near-term adoption, and government and private sector activity in these regions. The presentation provided a forecast of the global state of the industry in terms of shipments as well as a forecast of countries with emerging markets

329

Fuel production for LWRs - MOX fuel aspects  

International Nuclear Information System (INIS)

Plutonium recycling in Light Water Reactors is today an industrial reality. It is recycled in the form of (U, Pu)O2 fuel pellets (MOX), fabricated to a large extent according to UO2 technology and pellet design. The similarity of physical, chemical, and neutron properties of both fuels also allows MOX fuel to be burnt in nuclear plants originally designed to burn UO2. The industrial processes presently in use or planned are all based on a mechanical blending of UO2 and PuO2 powders. To obtain finely dispersed plutonium and to prevent high local concentration of plutonium, the feed materials are micronised. In the BNFL process, the whole (UO2, PuO2) blend is micronised by attrition milling. According to the MIMAS process, developed by BELGONUCLEAIRE, a primary blend made of UO2 containing about 30% PuO2 is micronised in a ball mill, afterwards this primary blend is mechanically diluted in UO2 to obtain the specified Pu content. After mixing, the (U, Pu)O2 powder is pressed and the pellets are sintered. The sintering cover gas contains moisture and 5 v/o H2. Moisture increases the sintering process and the U-Pu interdiffusion. After sintering and grinding, the pellets are submitted to severe controls to verify conformity with customer specifications (fissile content, Pu distribution, surface condition, chemical purity, density, microstructuon, chemical purity, density, microstructure). (author)

330

Fuel choice in new fossil fuel electric power plants  

Energy Technology Data Exchange (ETDEWEB)

The long-run choice of fuels in new fossil fuel electric power plants in the U.S. during 1955-1979 is analyzed. Econometric analyses are conducted of whether a plant is capable of using a particular fossil fuel, whether the fuel is actually used, and the input share of the fuel (given that the plant is capable of burning the fuel). Estimates of fuel substitution and price elasticities are presented. 16 refs. (A.V.)

Seifi, A. (Ministry of Energy, Teheran (Iran)); McDonald, J.F.

1986-03-01

331

GENUSA Fuel Evolution  

International Nuclear Information System (INIS)

GNF ENUSA Nuclear Fuel S.A. (GENUSA) was formed in Madrid in May 1996. GENUSA is a corporation organized and existing under the laws of Spain, jointly owned by GNF-A and ENUSA. GENUSA consolidates all European BWR fuel marketing activities of GNF-A and ENUSA, primarily providing marketing and project management. In its standard way of operating, it will obtain engineering, components and conversion from GNF-A and engineering, fabrication and fuel related services from ENUSA. GENUSA's development philosophy over the past decades has been to introduce evolutionary designs, supported by our global experience base, that deliver the performance needed by our customers to meet their operating strategies. GENUSA considers, as one of our strengths, the ever-increasing experience base that provides the foundation for such evolutionary changes. This experience is supported and complemented with an even greater GNF experience. Over the last 40 years, GNF and ENUSA have designed, fabricated, and placed in operation over 144,000 BWR fuel bundles containing over 9.7 million fuel rods. This experience base represents the widest range of operating conditions of any BWR fuel vendor, reflecting varying reactor power densities, operating strategies, and water chemistry environments. It covers operating periods of up to ?10 years and bundle average exposures up to 68 MWd/kgU.. It provides the confirmation of our understanding and ability to model fuel performance behavior, and has been fuel performance behavior, and has been instrumental in the identification and characterization of each encountered failure mechanism. With the knowledge gained from this extensive experience base, mitigating actions have been developed and progressively implemented by GENUSA as part of a continuous program toward improved fuel reliability and performance. GENUSA's evolutionary product introduction strategy has been extremely successful. There has been a continuous stream of new products/processes that were developed to deliver improved performance. Relative to the 8x8 fuel operated in the 1980's, today's designs provide ?25% more efficiency and power capability and twice as much energy. Because of GENUSA's evolutionary design commitment, these product improvements have been successfully rolled out to our customers with no design or fabrication-related performance surprises. Additionally, this has been accomplished with an accompanying steady improvement in fuel reliability. In the past three decades, fuel reliability has improved by approximately three orders of magnitude. That is, the fuel rod leaker rate has been reduced from over five hundred rods per million operating, to less than ten. In past decades, most plants experienced failures each cycle, and fleet-wide failure mechanisms drove reliability statistics. Today, a small minority of our customers' plants experience failures in any cycle, mainly recurrent, low level debris fretting failures in a handful of plants. GENUSA is committed to providing the most robust, and balanced, fuel solutions to our customers based on our extensive experience and technological capabilities. Identifying and successfully mitigating the mechanisms that cause fuel failures has been instrumental in this observed improvement in fuel reliability. GENUSA systematically identified and eliminated mechanisms leading to failure through pool-side and hot cell examinations, and feedback of lessons learned into the design and fabrication of the fuel. Some of the highly successful mitigating actions during this history include: - Improved pellet fabrication in the 1970's to eliminate cladding primary hydride failures; - Corrosion-resistant cladding, with a chemistry and microstructure specifically targeted to protect against crud-induced corrosion (CILC) failures; - Improved cladding and welding fabrication and inspection techniques that assured the hermeticity and quality of the delivered fuel rod; - Tightened pellet missing surface specifications to add PCI margin; - Introduction of a debris filter, applied as a standard feature

332

Fuel element development  

International Nuclear Information System (INIS)

In capsule irradiation tests the influence was studied which is exerted by high power densities on thin oxide fuel rods. Cladding expansions have been observed which are not attributable to creep but to plastic strains. Power jumps during load cycling resulted in stress to the cladding through fuel pressure due to thermal differential strain. - Changes in geometry of oxide fuel pellets during cycling were investigated theoretically using models. The test group 5b was also studied with a view to plutonium redistribution. A very high plutonium enrichment was found at the central channel, and outer zones nearly free from plutonium soon after the beginning of irradiation, which might be due to the high specific power and central temperature and the high PuO2-content (35%) of the fuel. Two contributions include as subjects the porosity of fuel in the context of structural analyses and creep caused by irradiation. The plutonium content itself does not seem to increase substantially the creep rate. Further results of post-examinations are available from the oxide irradiation tests Mol-7B and DFR-435. The zone of maximum damage of the Mol-7B-rods occurs at the upper end of the fuel column; even here the structure of the rod has essentially remained unchanged. The amount of fuel escaping is not as great as at the damaged points of DFR-435. (orig.)

333

Disposal of spent fuel  

International Nuclear Information System (INIS)

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

334

Integrated fuel management  

International Nuclear Information System (INIS)

At the Gentilly-2 NPGS, after an integration effort by the owner/operator Hydro-Quebec, the term FUEL MANAGEMENT has taken on a broad definition that encompasses all activities related to the complete fuel cycle. The activities include:planning, Q.A. and administration in the procurement of uranium and fuel bundles, out of core new/spent fuel inventories, in-core fuel management, analysis of fuel performance, and responding to AECB/AIEA accountancy requirement for international safeguards. This paper will describe how personnel involved in the above mentioned activities are organized under a single working unit for peak efficiency. It will also describe the integrated network of computerized systems used by this group. The system links on-site, corporate and outside data banks and programs, thus avoiding errors and the unnecessary, time consuming efforts involved in the duplication or manual input of required data to different computer codes. The decision to adopt integration as the way to better fuel management, and the commitment to pursue it as new ways and means come about, was never regretted and has contributed greatly to the excellent results obtained thus far at Gentilly-s in this area. 2 tabs

335

Bio-fuels barometer  

International Nuclear Information System (INIS)

European Union bio-fuel use for transport reached 12 million tonnes of oil equivalent (mtoe) threshold during 2009. The slowdown in the growth of European consumption deepened again. Bio-fuel used in transport only grew by 18.7% between 2008 and 2009, as against 30.3% between 2007 and 2008 and 41.8% between 2006 and 2007. The bio-fuel incorporation rate in all fuels used by transport in the E.U. is unlikely to pass 4% in 2009. We can note that: -) the proportion of bio-fuel in the German fuels market has plummeted since 2007: from 7.3% in 2007 to 5.5% in 2009; -) France stays on course with an incorporation rate of 6.25% in 2009; -) In Spain the incorporation rate reached 3.4% in 2009 while it was 1.9% in 2008. The European bio-diesel industry has had another tough year. European production only rose by 16.6% in 2009 or by about 9 million tonnes which is well below the previous year-on-year growth rate recorded (35.7%). France is leading the production of bio-ethanol fuels in Europe with an output of 1250 million liters in 2009 while the total European production reached 3700 million litters and the world production 74000 million liters. (A.C.)

336

Nuclear fuel element  

International Nuclear Information System (INIS)

In a case of pilling-up nuclear fuel pellets, deviation in the parallelness at the upper surface and the lower surface of individual pellets are accumulated or the axial direction is inclined on the unit of the pellet group. Particularly, the pellet end face with large deformation amount or deviation of parallelness gives large stress distortion to the fuel can. A fuel pellet formed with a concaved surface at a predetermined radius of curvature each of end faces and another fuel pellet formed with a convexed surface at a radius of curvature identical therewith at each of the end faces are combined with each other and charged in the fuel can. Thus, it is possible to eliminate the deviation for the parallelness of the pellets, prevent rattling of the pellets and reduce pellet-cladding mutual interaction. By forming the concaved fuel pellet with highly enriched uranium having a hollow portion while forming the convexed nuclear fuel pellet with less enriched uranium, heat generation at the central portion of the pellets can be suppressed. Accordingly, distortion of the pellet or release of FP gases can be reduced. (K.M.)

337

Alkaline fuel cells applications  

Science.gov (United States)

On the world-wide automobile market technical developments are increasingly determined by the dramatic restriction on emissions as well as the regimentation of fuel consumption by legislation. Therefore there is an increasing chance of a completely new technology breakthrough if it offers new opportunities, meeting the requirements of resource preservation and emission restrictions. Fuel cell technology offers the possibility to excel in today's motive power techniques in terms of environmental compatibility, consumer's profit, costs of maintenance and efficiency. The key question is economy. This will be decided by the costs of fuel cell systems if they are to be used as power generators for future electric vehicles. The alkaline hydrogen-air fuel cell system with circulating KOH electrolyte and low-cost catalysed carbon electrodes could be a promising alternative. Based on the experiences of Kordesch [K. Kordesch, Brennstoffbatterien, Springer, Wien, 1984, ISBN 3-387-81819-7; K. Kordesch, City car with H 2-air fuel cell and lead-battery, SAE Paper No. 719015, 6th IECEC, 1971], who operated a city car hybrid vehicle on public roads for 3 years in the early 1970s, improved air electrodes plus new variations of the bipolar stack assembly developed in Graz are investigated. Primary fuel choice will be a major issue until such time as cost-effective, on-board hydrogen storage is developed. Ammonia is an interesting option. The whole system, ammonia dissociator plus alkaline fuel cell (AFC), is characterised by a simple design and high efficiency.

Kordesch, Karl; Hacker, Viktor; Gsellmann, Josef; Cifrain, Martin; Faleschini, Gottfried; Enzinger, Peter; Fankhauser, Robert; Ortner, Markus; Muhr, Michael; Aronson, Robert R.

338

Hydrogen as automotive fuel  

International Nuclear Information System (INIS)

Hydrogen fueled vehicles may just be the answer to the air pollution problem in highly polluted urban environments where the innovative vehicle's air pollution abatement characteristics would justify its high operating costs as compared with those of conventional automotive alternatives. This paper examines the feasibility of hydrogen as an automotive fuel by analyzing the following aspects: the chemical-physical properties of hydrogen in relation to its use in internal combustion engines; the modifications necessary to adapt internal combustion engines to hydrogen use; hydrogen fuel injection systems; current production technologies and commercialization status of hydrogen automotive fuels; energy efficiency ratings; environmental impacts; in-vehicle storage systems - involving the use of hydrides, high pressure systems and liquid hydrogen storage systems; performance in terms of pay-load ratio; autonomous operation; and operating costs. With reference to recent trial results being obtained in the USA, an assessment is also made of the feasibility of the use of methane-hydrogen mixtures as automotive fuels. The paper concludes with a review of progress being made by ENEA (the Italian Agency for New Technology, Energy and the Environment) in the development of fuel storage and electronic fuel injection systems for hydrogen powered vehicles

339

Nuclear fuel assembly  

International Nuclear Information System (INIS)

For preventing the burn-out of fuel rods and increasing the operatioin margin of a nuclear reactor, cooling performance at the upper portion where PIP (Power Integral Power) is high has to be increased more. In view of the above, in the nuclear fuel assembly according to the present invention partial fins having different length and axial angle and plate-like spacers in parallel with the axial line are disposed at the outer circumference of a fuel rod at axially uneven pitch, in which the length and the angle of inclination of the partial fins are varied depending on the positions. Further, fuel rods in adjacent with each other are maintained axially slidably so as to form a bundle of fuel rods while maintaining the gap therebetween. This can optimize the distribution of the cooling performance in the longitudinal direction of the nuclear fuel assembly and enables to form a tubular bundle at a low pressure loss. Further, since the size of the spacer can be reduced, it is applicable also to a case where density of the fuel rod arrangement is rather increased and fretting abrasion can also be decreased. (T.M.)

340

Gamma absorption fuel densitometer  

International Nuclear Information System (INIS)

A gamma absorption fuel densitometer for measuring the fissile material density and uniformity in a fuel pin based on transmission studies has been set up. Cobalt-57 was used as the gamma ray source, a bench mechanism was used to achieve uniform motion of the fuel pin and a well-shielded and collimated NaI (Tl) detector was used to measure the transmitted gamma rays. The feasibility studies indicated that the density distribution in the pin and its active length could be reliably measured using this system. (author)

341

Fuel rod technology  

International Nuclear Information System (INIS)

By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.)

342

Nuclear fuel spacer grid  

International Nuclear Information System (INIS)

A typical embodiment of the invention provides structural support for the grids in a nuclear reactor fuel assembly. Illustratively, the external surfaces of water rods or instrument tubes in a fuel assembly are provided with annular recesses. A spacer grid retainer is engaged within the recesses by means of spring loaded fingers on the retainer, in which the spring loading forces are greater than anticipated vibration forces to reduce fretting corrosion. Notches formed in the retainer secure the grid to the retainer, all in a manner that simplifies fuel assembly construction and restricts grid movement at lower cost

343

Nuclear fuel elements  

International Nuclear Information System (INIS)

Object: To effectively occlude injurious substances such as water contents and hydrogen gas present in the nuclear fuel rod and to occlude excessive water contents, and thereafter to prevent inevitable discharge of the hydrogen gas. Structure: A nuclear fuel element having a cladding pipe in which uranium fuel pellets are received, the element sealing therein zirconium-titanium-nickel system alloy powder of 60 to 100 thickness and 42 and 20 mesh, and zirconium-titanium-nickel system alloy massive particles of 0.1 to 0.5 thickness. (Kamimura, M.)

344

Nuclear fuel element  

International Nuclear Information System (INIS)

Purpose: To reduce the local stress acting on a cladding tube by forming a plurality of heat resistant projections on the outer circumferential surface of a fuel pellet. Constitution: Upon thermal expansion of fuel pellets during use, only the heat resistant projections formed by flame spray coating on the outer circumferential surface thereof are contacted to the inner surface of the cladding tube, so that if cracks are resulted in the fuel pellet, no excessive local tensile strength is exerted on the portion of the cladding tube corresponding to the cracks. (Seki, T.)

345

Fuel storage tank  

International Nuclear Information System (INIS)

The stationary fuel storage tank is immersed below the water level in the spent fuel storage pool. In it there is placed a fuel assembly within a cage. Moreover, the storage tank has got a water filling and a gas buffer. The water in the storage tank is connected with the pool water by means of a filter, a surge tank and a water purification facility, temperature and pressure monitoring being performed. In the buffer compartment there are arranged catalysts a glow plugs for recombination of radiolysis products into water. The supply of water into the storage tank is performed through the gas buffer compartment. (DG)

346

GENUSA Fuel Evolution  

Energy Technology Data Exchange (ETDEWEB)

GNF ENUSA Nuclear Fuel S.A. (GENUSA) was formed in Madrid in May 1996. GENUSA is a corporation organized and existing under the laws of Spain, jointly owned by GNF-A and ENUSA. GENUSA consolidates all European BWR fuel marketing activities of GNF-A and ENUSA, primarily providing marketing and project management. In its standard way of operating, it will obtain engineering, components and conversion from GNF-A and engineering, fabrication and fuel related services from ENUSA. GENUSA's development philosophy over the past decades has been to introduce evolutionary designs, supported by our global experience base, that deliver the performance needed by our customers to meet their operating strategies. GENUSA considers, as one of our strengths, the ever-increasing experience base that provides the foundation for such evolutionary changes. This experience is supported and complemented with an even greater GNF experience. Over the last 40 years, GNF and ENUSA have designed, fabricated, and placed in operation over 144,000 BWR fuel bundles containing over 9.7 million fuel rods. This experience base represents the widest range of operating conditions of any BWR fuel vendor, reflecting varying reactor power densities, operating strategies, and water chemistry environments. It covers operating periods of up to {approx}10 years and bundle average exposures up to 68 MWd/kgU.. It provides the confirmation of our understanding and ability to model fuel performance behavior, and has been instrumental in the identification and characterization of each encountered failure mechanism. With the knowledge gained from this extensive experience base, mitigating actions have been developed and progressively implemented by GENUSA as part of a continuous program toward improved fuel reliability and performance. GENUSA's evolutionary product introduction strategy has been extremely successful. There has been a continuous stream of new products/processes that were developed to deliver improved performance. Relative to the 8x8 fuel operated in the 1980's, today's designs provide {approx}25% more efficiency and power capability and twice as much energy. Because of GENUSA's evolutionary design commitment, these product improvements have been successfully rolled out to our customers with no design or fabrication-related performance surprises. Additionally, this has been accomplished with an accompanying steady improvement in fuel reliability. In the past three decades, fuel reliability has improved by approximately three orders of magnitude. That is, the fuel rod leaker rate has been reduced from over five hundred rods per million operating, to less than ten. In past decades, most plants experienced failures each cycle, and fleet-wide failure mechanisms drove reliability statistics. Today, a small minority of our customers' plants experience failures in any cycle, mainly recurrent, low level debris fretting failures in a handful of plants. GENUSA is committed to providing the most robust, and balanced, fuel solutions to our customers based on our extensive experience and technological capabilities. Identifying and successfully mitigating the mechanisms that cause fuel failures has been instrumental in this observed improvement in fuel reliability. GENUSA systematically identified and eliminated mechanisms leading to failure through pool-side and hot cell examinations, and feedback of lessons learned into the design and fabrication of the fuel. Some of the highly successful mitigating actions during this history include: - Improved pellet fabrication in the 1970's to eliminate cladding primary hydride failures; - Corrosion-resistant cladding, with a chemistry and microstructure specifically targeted to protect against crud-induced corrosion (CILC) failures; - Improved cladding and welding fabrication and inspection techniques that assured the hermeticity and quality of the delivered fuel rod; - Tightened pellet missing surface specifications to add PCI margin; - Introduction of a debris filter, applied as a standard fe

Choithramani, Sylvia; Malpica, Maria [ENUSA Industrias Avanzadas, GENUSA, Josefa Valcarcel, 26 28027 Madrid (Spain); Fawcett, Russel [Global Nuclear Fuel (United States)

2009-06-15

347

The Dounreay fuel cycle facilities of AEA Fuel Services  

International Nuclear Information System (INIS)

Having developed the fuel technology for Britain's fast reactor programme, AEA Technology's Dounreay site now boasts one of the most flexible fuel cycle facilities in the world. In this article, these facilities and the many services that AEA Fuel Services provide to the specialist fuel market are described from fuel development to fabrication and reprocessing. (author)

348

Nuclear fuel waste disposal  

International Nuclear Information System (INIS)

This report discusses events and processes that could adversely affect the long-term stability of a nuclear fuel waste disposal vault or the regions of the geosphere and the biosphere to which radionuclides might migrate from such a vault

349

North Korea's corroding fuel  

International Nuclear Information System (INIS)

The roughly 8,000 irradiated or open-quotes spentclose quotes fuel rods recently discharged from the North Korean 25 megawatt (thermal) reactor are difficult to store safely under the conditions in the spent fuel ponds near the reactor. The magnesium alloy jacket, or open-quotes cladding,close quotes around the fuel elements is corroding. If the corrosion creates holes in the cladding, radionuclides may be released. In addition, the uranium metal underneath the cladding may begin to corrode, possibly creating uranium hydride which can spontaneously ignite in air. Unless the storage conditions are improved, North Korea may use the risk posed by the corrosion as an argument for reprocessing this fuel, a violation of its June 1994 pledge to the United States to freeze its nuclear program. North Korea, however, can take several steps to slow dramatically the rate of corrosion. Using available techniques, it can extend safe storage times by months or even years

350

Solid Oxide Fuel Cell  

International Science & Technology Center (ISTC)

Development and Demonstration of the Advance Technology of New Type Production of Pipe High Temperature Solid Oxide Fuel Cells and Making of Pilot Samples of these Elements for Standard Conditions of Application in Electrochemical Current Sources

351

Nuclear fuel cycle techniques  

International Nuclear Information System (INIS)

The production of fuels for nuclear power plants involves five principal stages: prospecting of uranium deposits (on the ground, aerial, geochemical, geophysical, etc...); extraction and production of natural uranium from the deposits (U content of ores is not generally high and a chemical processing is necessary to obtain U concentrates); production of 235U enriched uranium for plants utilizing this type of fuel (a description is given of the gaseous diffusion process widely used throughout the world and particularly in France); manufacture of suitable fuel elements for the different plants; reprocessing of spent fuels for the purpose of not only recovering the fissile materials but also disposing safely of the fission products and other wastes

352

Renewable jet fuel.  

Science.gov (United States)

Novel strategies for sustainable replacement of finite fossil fuels are intensely pursued in fundamental research, applied science and industry. In the case of jet fuels used in gas-turbine engine aircrafts, the production and use of synthetic bio-derived kerosenes are advancing rapidly. Microbial biotechnology could potentially also be used to complement the renewable production of jet fuel, as demonstrated by the production of bioethanol and biodiesel for piston engine vehicles. Engineered microbial biosynthesis of medium chain length alkanes, which constitute the major fraction of petroleum-based jet fuels, was recently demonstrated. Although efficiencies currently are far from that needed for commercial application, this discovery has spurred research towards future production platforms using both fermentative and direct photobiological routes. PMID:24679258

Kallio, Pauli; Pásztor, András; Akhtar, M Kalim; Jones, Patrik R

2014-04-01

353

Fuel element pond  

International Nuclear Information System (INIS)

The fuel element pond is part of a fuel reprocessing centre, to which spent fuel elements from LWR's are supplied, for example. The fuel element transport containers are driven to and fro in the dry state. During loading and unloading of transport containers, the pond is filled with water. After emptying the pond again, its walls may be contaminated. Cleaning of the pond walls takes place with a spray device, which is remotely operated. It consits of a pipe, which is connected to the cleaning liquid (water with citric acid) and which has spray nozzles. It may be connected to a float so that it follows the different water levels, or the nozzles can be closed according to the water level. (DG)

354

Secondary fuel delivery system  

Science.gov (United States)

A secondary fuel delivery system for delivering a secondary stream of fuel and/or diluent to a secondary combustion zone located in the transition piece of a combustion engine, downstream of the engine primary combustion region is disclosed. The system includes a manifold formed integral to, and surrounding a portion of, the transition piece, a manifold inlet port, and a collection of injection nozzles. A flowsleeve augments fuel/diluent flow velocity and improves the system cooling effectiveness. Passive cooling elements, including effusion cooling holes located within the transition boundary and thermal-stress-dissipating gaps that resist thermal stress accumulation, provide supplemental heat dissipation in key areas. The system delivers a secondary fuel/diluent mixture to a secondary combustion zone located along the length of the transition piece, while reducing the impact of elevated vibration levels found within the transition piece and avoiding the heat dissipation difficulties often associated with traditional vibration reduction methods.

Parker, David M. (Oviedo, FL); Cai, Weidong (Oviedo, FL); Garan, Daniel W. (Orlando, FL); Harris, Arthur J. (Orlando, FL)

2010-02-23

355

Packing Nuclear Fuel  

International Science & Technology Center (ISTC)

Development of Scientific Foundations of the Technology of the Metal Matrix Packing of Leaky Unreprocessible Spent Nuclear Fuel of Different Purpose Reactors for a Long-term Environmentally Safe Storage.

356

Wastes from fuel reprocessing  

International Nuclear Information System (INIS)

Handling, treatment, and interim storage of radioactive waste, problems confronted with during the reprocessing of spent fuel elements from LWR's according to the Purex-type process, are dealt with in detail. (HR/LN)

357

Structure of Fuel Elements  

International Science & Technology Center (ISTC)

Study of Structure of Materials Based on U-Pu-Zr alloy, their Thermodynamic Properties and Interaction with Materials of Fuel Element Shell under Quasi-Isothermal Conditions and Conditions of Non-stationary Exposure

358

Nuclear fuel element  

International Nuclear Information System (INIS)

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

359

Ceramic nuclear fuel pellets  

International Nuclear Information System (INIS)

Low density nuclear fuel pellets are produced by mixing uranium dioxide powder and/or plutonium dioxide powder with ammonium oxalate, forming the mixture into pellets and sintering the pellets. 1 claim, 2 figures

360

Fuel cycle studies  

International Nuclear Information System (INIS)

Programs are being conducted in the following areas: advanced solvent extraction techniques, accident consequences, fuel cycles for nonproliferation, pyrochemical and dry processes, waste encapsulation, radionuclide transport in geologic media, hull treatment, and analytical support for LWBR

361

Hydrogen as a fuel  

Energy Technology Data Exchange (ETDEWEB)

A panel of the Committee on Advanced Energy Storage Systems of the Assembly of Engineering has examined the status and problems of hydrogen manufacturing methods, hydrogen transmission and distribution networks, and hydrogen storage systems. This examination, culminating at a time when rapidly changing conditions are having noticeable impact on fuel and energy availability and prices, was undertaken with a view to determining suitable criteria for establishing the pace, timing, and technical content of appropriate federally sponsored hydrogen R and D programs. The increasing urgency to develop new sources and forms of fuel and energy may well impact on the scale and timing of potential future hydrogen uses. The findings of the panel are presented. Chapters are devoted to hydrogen sources, hydrogen as a feedstock, hydrogen transport and storage, hydrogen as a heating fuel, automotive uses of hydrogen, aircraft use of hydrogen, the fuel cell in hydrogen energy systems, hydrogen research and development evaluation, and international hydrogen programs.

1979-01-01

362

Nuclear fuel assembly  

International Nuclear Information System (INIS)

Object: To increase the efficiency of core spray cooling used in case of an emergency accident such as an accident of loss of coolant. Structure: In a fuel assembly of a BWR type nuclear reactor, the upper tie plate for securing the fuel rod is constructed as a double-stage structure, and the fuel rods are made to have different lengths between those in the central portion and those in the edge portion. With the flow path formed in the upper tie plate the interaction between the vapor generated in the inside of the fuel assembly and core spray water from above is reduced, permitting both the fluids to smoothly flow in and out and thus increasing the efficiency of core spray cooling. (Horiuchi, T.)

363

Spent fuel storage pool  

International Nuclear Information System (INIS)

Purpose: To increase the capacities of spent fuel pools in the reactor building by dividing a pair, left and right, of pools communicating with the reactor well pool into two, upper and lower, stages and storing spent fuels in the lower stage. Constitution: At the lower part of a pool communicating with the reactor well pool through pool gates and shield blocks there are provided storage racks thereby to use the pool as an auxiliary machine tentatively set pool concurrently used as a spent fuel storage pool. Furthermore, a pair of spent fuel storage pools are provided at the left and right hand sides of the well pool thereby to double the storage capacity. (Sekiya, K.)

364

Fuel cell cogeneration  

Energy Technology Data Exchange (ETDEWEB)

The U.S. Department of Energy`s Morgantown Energy Technology Center (METC) sponsors the research and development of engineered systems which utilize domestic fuel supplies while achieving high standards of efficiency, economy, and environmental performance. Fuel cell systems are among the promising electric power generation systems that METC is currently developing. Buildings account for 36 percent of U.S. primary energy consumption. Cogeneration systems for commercial buildings represent an early market opportunity for fuel cells. Seventeen percent of all commercial buildings are office buildings, and large office buildings are projected to be one of the biggest, fastest-growing sectors in the commercial building cogeneration market. The main objective of this study is to explore the early market opportunity for fuel cells in large office buildings and determine the conditions in which they can compete with alternative systems. Some preliminary results and conclusions are presented, although the study is still in progress.

Wimer, J.G. [Dept. of Energy, Morgantown, WV (United States); Archer, D.

1995-08-01

365

Optical fuel spray measurements  

Energy Technology Data Exchange (ETDEWEB)

Diesel fuel sprays, including fuel/air mixing and the physics of two-phase jet formation, are discussed in the thesis. The fuel/air mixing strongly affects emissions formation in spray combustion processes where the local combustion conditions dictate the emission formation. This study comprises optical measurements both in pressurized spray test rigs and in a running engine.The studied fuel injection was arranged with a common rail injection system and the injectors were operated with a solenoid-based injection valve. Both marine and heavy-duty diesel engine injectors were used in the study. Optical fuel spray measurements were carried out with a laser-based double-framing camera system. This kind of equipments is usually used for flow field measurements with Particle Image Velocimetry technique (PIV) as well as for backlight imaging. Fundamental fuel spray properties and spray formation were studied in spray test rigs. These measurements involved studies of mixing, atomization, and the flow field. Test rig measurements were used to study the effect of individual injection parameters and component designs. Measurements of the fuel spray flow field, spray penetration, spray tip velocity, spray angle, spray structure, droplet accumulation, and droplet size estimates are shown. Measurement campaign in a running optically accessible large-bore medium-speed engine was also carried out. The results from engine tests were compared with equivalent test rig measurements, as well as computational results, to evaluate the level of understanding of sprays. It was shown that transient spray has an acceleration and a deceleration phase. Successive flow field measurements (PIV) in optically dense diesel spray resulted in local and average velocity data of diesel sprays. Processing fuel spray generates a flow field to surrounding gas and entrainment of surrounding gas into fuel jet was also seen at the sides of the spray. Laser sheet imaging revealed the inner structure of diesel spray and accumulation of droplets. Also shockwave formation was recorded when supersonic fuel jet exits the nozzle orifice. These results were used to evaluate spray formation and the structure was compared with simulated fuel sprays. Novel information, more refined and focused results, and better understanding of the nature of atomization and sprays was gathered. It was shown that new methods enable more precise understanding of transient two-phase sprays to be gained. (orig.)

Hillamo, H.

2011-07-01

366

IFR fuel cycle  

International Nuclear Information System (INIS)

The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase, which includes completion of facility modifications and installation and cold checkout of process equipment. This paper reviews the development of the electrorefining pyroprocess, the design and construction of the facility for the hot demonstration, the design and fabrication of the equipment, and the schedule and initial plan for its operation

367

USCEA fuel cycle '93  

International Nuclear Information System (INIS)

The US Council for Energy Awareness sponsored the Fuel Cycle '93 conference in Dallas, Texas, on March 21-24, 1993. Over 250 participants attended, numerous papers were presented, and several panel discussions were held. The focus of most industry participants remains the formation of USEC and the pending US-Russian HEU agreement. Following are brief summaries of two key papers and the Fuel Market Issues panel discussion

368

Fuel reprocessing world prospective  

International Nuclear Information System (INIS)

Reprocessing should be considered as the reference solution for spent fuel management, as it allows recovering and recycling of reusable materials, uranium and plutonium, and the reduction of ultimate residues at the lowest level possible. Plutonium production and proliferation issues may be solved through the use of MOX fuels, which enables a zero (or even negative) plutonium balance. Regulations concerning ultimate wastes, whether they will authorize maximum levels at 1% or 0.1% of plutonium, will influence reprocessing future prospects

369

Composite fuel cell membranes  

Science.gov (United States)

A bilayer or trilayer composite ion exchange membrane suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

Plowman, Keith R. (Lake Jackson, TX); Rehg, Timothy J. (Lake Jackson, TX); Davis, Larry W. (West Columbia, TX); Carl, William P. (Marble Falls, TX); Cisar, Alan J. (Cypress, TX); Eastland, Charles S. (West Columbia, TX)

1997-01-01

370

Fast breeder fuel cycle  

International Nuclear Information System (INIS)

Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

371

Hydrogen: Fueling the Future  

International Nuclear Information System (INIS)

As our dependence on foreign oil increases and concerns about global climate change rise, the need to develop sustainable energy technologies is becoming increasingly significant. Worldwide energy consumption is expected to double by the year 2050, as will carbon emissions along with it. This increase in emissions is a product of an ever-increasing demand for energy, and a corresponding rise in the combustion of carbon containing fossil fuels such as coal, petroleum, and natural gas. Undisputable scientific evidence indicates significant changes in the global climate have occurred in recent years. Impacts of climate change and the resulting atmospheric warming are extensive, and know no political or geographic boundaries. These far-reaching effects will be manifested as environmental, economic, socioeconomic, and geopolitical issues. Offsetting the projected increase in fossil energy use with renewable energy production will require large increases in renewable energy systems, as well as the ability to store and transport clean domestic fuels. Storage and transport of electricity generated from intermittent resources such as wind and solar is central to the widespread use of renewable energy technologies. Hydrogen created from water electrolysis is an option for energy storage and transport, and represents a pollution-free source of fuel when generated using renewable electricity. The conversion of chemical to electrical energy using fuel cells provides a high efficiey using fuel cells provides a high efficiency, carbon-free power source. Hydrogen serves to blur the line between stationary and mobile power applications, as it can be used as both a transportation fuel and for stationary electricity generation, with the possibility of a distributed generation energy infrastructure. Hydrogen and fuel cell technologies will be presented as possible pollution-free solutions to present and future energy concerns. Recent hydrogen-related research at SLAC in hydrogen production, fuel cell catalysis, and hydrogen storage will be highlighted in this seminar.

372

Nuclear reactor fuel assemblies  

International Nuclear Information System (INIS)

A nuclear reactor fuel stringer comprises at least two generally co-extensive and generally rectilinear tie elements which extend centrally through the stringer and are linked in such a way that one or more of the elements can support the fuel stringer load. In one embodiment there are two coaxial tie elements. In another embodiment six tie elements are arranged in a circle. (Author)

373

Scintillator spent fuel monitor  

International Nuclear Information System (INIS)

A monitor for rapidly measuring the gross gamma-ray flux immediately above spent fuel assemblies in underwater storage racks has been developed. It consists of a plastic scintillator, photomultiplier, collimator, and a small battery-powered electronics package. The crosstalk from an isolated fuel assembly to an adjacent void is only about 2%. The mean difference between the measured gamma-ray flux and the flux estimated from the declared burnup and cooling time with a simple formula is 22%

374

Nuclear fuel elements  

International Nuclear Information System (INIS)

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

375

Fuel Cell History  

Science.gov (United States)

This paper from George Wand outlines the history of fuel cells and hydrogen use, beginning with historical information on battery powered electric vehicles and moving through the decades and development of a variety of different vehicles. The end of the report takes a brief look into the possible future of hydrogen and fuel cell technologies to power automobiles. This document may be downloaded in PDF file format.

Wand, George

376

Compliant fuel cell system  

Science.gov (United States)

A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

Bourgeois, Richard Scott (Albany, NY); Gudlavalleti, Sauri (Albany, NY)

2009-12-15

377

Fuels for Transportation  

OpenAIRE

There is a need to reduce the amount of fossil energy used for transport, both because of the easily available fossil fuel is becoming sparser and because of climate concerns. In this article, the concept of “peak oil” is briefly presented. Second, a practical approach to reduction of fossil fuel use for transport elaborated by two British commissions is presented. A key feature is the introduction of electric cars. This raises the third issue covered in this article: namely, how battery ...

Fredholm, Bertil B.; Norde?n, Bengt

2010-01-01

378

Vibrating fuel grapple  

Science.gov (United States)

A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

Chertock, deceased, Alan J. (late of San Francisco, CA); Fox, Jack N. (San Jose, CA); Weissinger, Robert B. (Santa Clara, CA)

1982-01-01

379

Automatic fuel exchanger  

International Nuclear Information System (INIS)

Purpose: To previously detect the difference in the insertion angle caused by the erroneous positioning of a cask to an aimed coordinate or the difference in the insertion angle resulted from the abnormality in the rotational system thereby preventing the lowering of a catcher or insertion of fuel at such incorrect angle to avoid the damages to the equipments. Constitution: A catcher is set at an aimed angle by actuating a monitor from a catcher rotation control section and the catcher is moved vertically from another motor to insert or extract fuel rods. In this case, detection means (image sensor or the like) is disposed for detecting the fuel loading state in the reactor core or the fuel pool as quantitized images in a non-contact manner by utilizing photoelectronic conversion techniques. A quantitized image processing section processes the quantitized images from the sensor to judge the insertion angle of the fuel rod. A comparison section compares the coincidence among the judged insertion angle, the insertion angle of the fuel catcher determined from detectors and a present catcher angle judging section and a catcher rotation instruction relative to the aimed coordinate and issues an insertion inhibitive instruction in case they do not coincide. (Horiuchi, T.)

380

Thorium fuel cycles  

International Nuclear Information System (INIS)

Almost every type of reactor has been associated at one time or another with a proposal to utilize a thorium fuel cycle. Commercial-scale experience with the use of thorium fuel cycles has however been extremely limited to date. Thorium cycles offer the attraction of good fissile material utilization in thermal reactors, thorium could in principle have commercial attractions in this application. However, the U-Pu fast reactor cycle, has even better fissile material utilization. Thorium is inferior to depleted uranium as a fertile material in fast reactors. The article considers in detail the use of thorium fuel cycles both in thermal and fast reactors, and the parameters governing the choice between thorium-based and uranium-based cycles in these various applications. All stages of a thorium fuel cycle, including the mining of ore, conversion, and the reprocessing and fabrication of 233U fuels must be taken into account when assessing its merits. In the unlikely event that thorium reprocessing were available, while for some reason 238U-plutonium reprocessing were not, then the HTR, HWR and LWR would all be possible candidates for thorium-based thermal reactor fuel cycles. (author)

381

Centralised spent fuel storage  

International Nuclear Information System (INIS)

France has a large nuclear power programme with about 50 nuclear power plants generating roughly 1300t of spent fuel a year. She has decided to continue reprocessing and recycling the recovered plutonium. This is being done for LWR fuels at La Hague, where the existing plant has been already carried out prompt reprocessing of more than 1300t. Two additional large 800tU/year plants, UP3 and UP2 800, will be on line in 1988 and 1991 respectively, to accommodate domestic fuel and meet the needs of foreign clients. To receive these fuels, a major spent fuel reception and storage programme has been implemented at La Hague. This facility is described and each stage from reception and unloading and storage in pools is explained. At La Hague there is dry reception for unloading the fuels into the new pools. For smaller storage capacities, dry storage is cheaper but for large capacities the pool is less expensive. Loading and unloading can be automated. Centralised storage prior to reprocessing reduces the amount of handling and reshuffling required. (UK)

382

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To facilitate the production facility and eliminate the requirement for the long storage of fuels after the use as the radioactive wastes, by incorporating burnable poisons into a fuel rod in a separated structure while not being mixed with UO2 pellets. Method: A fuel rod for use in water cooled reactors comprises an appropriate number of cylindrical pellets made of nuclear fuel material charged in a fuel can and then tightly shielded with upper and lower end plugs. The axial portion of the cylindrical pellets is made hollow, through which an elongate rod incorporated with burnable poisons is charged vertically. The poisons can control the excess reactivity of the reactor core and flatten the neutron flux distribution. While sintered pellet rods of boron 10 were inserted to the gap of the fuel assembly or gadlinia was admixture with uranium dioxide, the former has been dissipated during use and the latter complicates the production step due to the residue after the reprocession. These problems can be dissolved by this invention. (Kamimura, M.)

383

Closing the fuel cycle  

International Nuclear Information System (INIS)

The progressive implementation of some key nuclear fuel cycle capecities in a country corresponds to a strategy for the acquisition of an independant energy source, France, Japan, and some European countries are engaged in such strategic programs. In France, COGEMA, the nuclear fuel company, has now completed the industrial demonstration of the closed fuel cycle. Its experience covers every step of the front-end and of the back-end: transportation of spent fuels, storage, reprocessing, wastes conditioning. The La Hague reprocessing plant smooth operation, as well as the large investment program under active progress can testify of full mastering of this industry. Together with other French and European companies, COGEMA is engaged in the recycling industry, both for uranium through conversion of uranyl nitrate for its further reeichment, and for plutonium through MOX fuel fabrication. Reprocessing and recycling offer the optimum solution for a complete, economic, safe and future-oriented fuel cycle, hence contributing to the necessary development of nuclear energy. (author)

384

Fuel cladding tube  

International Nuclear Information System (INIS)

Zirconium liner is formed on the inner surface of a base material of a fuel cladding tube made of a zirconium alloy (for example, zircaloy-2). A low purity zirconium (sponge zirconium) is used as the zirconium liner. The low purity zirconium contains 600ppm of Fe and 500ppm of oxygen as main impurities. The crystals of low purity zirconium constituting the zirconium liner have a fr value of 0.65 or greater in the radial direction of the fuel cladding tube. The fr value is a relative ratio of a c-axis of crystals oriented in parallel with the radial direction of fuel cladding tube. Since a drawing fabrication is applied to the fuel cladding tube so as to concentrate the c-axis of the crystals to the radial direction of the fuel cladding tube, a possibility of stress corrosion cracks is lowered. Accordingly, a fuel cladding tube having high reliability with less failure possibility can be attained. (I.N.)

385

Nuclear fuel assembly  

International Nuclear Information System (INIS)

Purpose: To improve the reactor stability by increasing the margin to fuel rod failures in a nuclear fuel assemblies of BWR type reactors, as well as reducing the pressure drop in the assemblies. Constitution: Seven spacers are disposed at an equi-distance between effective heat generating portions of a fuel rod. Among these spacers, the structure for the spacers disposed to the downstream of the coolant flow is made different from that of other spacers. That is, protrusions are disposed to the outermost circum-ference of the spacers to convert the flow of the uprising coolants downwardly such that liquid droplets in the circular vapor flow are readily deposited to the surface of the fuel rod. As a result, the thickness of the liquid films at the surface of the fuel rod is increased to suppress transition boiling. Further, since no protrusions are disposed to other spacers, the pressure drop in the entire fuel assembly is reduced to improve the reactor core stability. (Kamimura, M.)

386

Fuel exchanger control device  

International Nuclear Information System (INIS)

Purpose: To improve the stability and the operationability of the fuel exchanging work by checking the validity of the data before the initiation of the work. Constitution: A floppy disc stores the initial charging state data showing the arrangement of fuel assemblies in the reactor core pool, data showing the working procedures for the fuel exchange and a final charged state data upon completion of the work. The initial data and the procedure data are read from the disk and stored once into a memory. Then, the initial data are sequentially performed on the memory in accordance with the procedure data and, thereafter, they were compared with the final data read from the disk. After confirming that there are no errors in the working data, the procedure data are orderly instructed to the fuel exchanger for performing fuel replacement. Accordingly, since the data are checked before the initiation of the work, the fuel exchange can be performed automatically thereby improving the operationability thereof. (Yoshino, Y.)

387

Alternative Fuels: Research Progress  

Directory of Open Access Journals (Sweden)

Full Text Available Chapter 1: Pollutant Emissions and Combustion Characteristics of Biofuels and Biofuel/Diesel Blends in Laminar and Turbulent Gas Jet Flames. R. N. Parthasarathy, S. R. Gollahalli Chapter 2: Sustainable Routes for The Production of Oxygenated High-Energy Density Biofuels from Lignocellulosic Biomass. Juan A. Melero, Jose Iglesias, Gabriel Morales, Marta Paniagua Chapter 3: Optical Investigations of Alternative-Fuel Combustion in an HSDI Diesel Engine. T. Huelser, M. Jakob, G. Gruenefeld, P. Adomeit, S. Pischinger Chapter 4: An Insight into Biodiesel Physico-Chemical Properties and Exhaust Emissions Based on Statistical Elaboration of Experimental Data. Evangelos G. Giakoumis Chapter 5: Biodiesel: A Promising Alternative Energy Resource. A.E. Atabani Chapter 6: Alternative Fuels for Internal Combustion Engines: An Overview of the Current Research. Ahmed A. Taha, Tarek M. Abdel-Salam, Madhu Vellakal Chapter 7: Investigating the Hydrogen-Natural Gas Blends as a Fuel in Internal Combustion Engine. ?lker YILMAZ Chapter 8: Conversion of Bus Diesel Engine into LPG Gaseous Engine; Method and Experiments Validation. M. A. Jemni , G. Kantchev , Z. Driss , R. Saaidia , M. S. Abid Chapter 9: Predicting the Combustion Performance of Different Vegetable Oils-Derived Biodiesel Fuels. Qing Shu, ChangLin Yu Chapter 10: Production of Gasoline, Naphtha, Kerosene, Diesel, and Fuel Oil Range Fuels from Polypropylene and Polystyrene Waste Plastics Mixture by Two-Stage Catalytic Degradation using ZnO. Moinuddin Sarker, Mohammad Mamunor Rashid

Maher A.R. Sadiq Al-Baghdadi

2013-01-01

388

Fuel cleanup systems for fusion fuel processing  

International Nuclear Information System (INIS)

The fuel cleanup unit (FCU) of a fusion fuel processing system receives gas from the torus evacuation system which is composed of hydrogen isotopes (deuterium and tritium) contaminated with a few to possibly around 20% of impurities (methane, helium, protium, water, nitrogen, carbon oxides, ammonia, etc.). Because the usual processing step subsequent to the FCU, the cryogenic isotope separation system (ISS), operates at about 25 K, the FCU must remove all impurities which freeze at this temperature to prevent ISS plugging. This includes all impurities other than helium and protium. This first function of the FCU is termed purification. The collected impurities contain a substantial quality of tritium largely in the form of methane (and higher hydrocarbons), water and ammonia. This tritium must be recovered from the impurities before the remaining non-radioactive components may be discarded. This second function is referred to as tritium recovery. To accomplish these two functions a number of techniques have been proposed and investigated. This paper presents these techniques and identifies the merits and range of utility of each method. (orig.)

389

Fuel performance, design and development  

International Nuclear Information System (INIS)

The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

390

Failed fuel detection from viewpoint of fuel inspection  

International Nuclear Information System (INIS)

The cause of the failed fuel problem is discussed from the viewpoint of the inspection in fuel processing. Especially the problem of (1) the qualification of cladding and (2) the qualification of fuel rod welding are mentioned in detail. (auth.)

391

Solid oxide fuel cell generator  

Science.gov (United States)

A solid oxide fuel cell generator has a plenum containing at least two rows of spaced apart, annular, axially elongated fuel cells. An electrical conductor extending between adjacent rows of fuel cells connects the fuel cells of one row in parallel with each other and in series with the fuel cells of the adjacent row.

Di Croce, A. Michael (Murrysville, PA); Draper, Robert (Churchill Boro, PA)

1993-11-02

392

Hydrogen-enriched fuels  

Energy Technology Data Exchange (ETDEWEB)

NRG Technologies, Inc. is attempting to develop hardware and infrastructure that will allow mixtures of hydrogen and conventional fuels to become viable alternatives to conventional fuels alone. This commercialization can be successful if the authors are able to achieve exhaust emission levels of less than 0.03 g/kw-hr NOx and CO; and 0.15 g/kw-hr NMHC at full engine power without the use of exhaust catalysts. The major barriers to achieving these goals are that the lean burn regimes required to meet exhaust emissions goals reduce engine output substantially and tend to exhibit higher-than-normal total hydrocarbon emissions. Also, hydrogen addition to conventional fuels increases fuel cost, and reduces both vehicle range and engine output power. Maintaining low emissions during transient driving cycles has not been demonstrated. A three year test plan has been developed to perform the investigations into the issues described above. During this initial year of funding research has progressed in the following areas: (a) a cost effective single-cylinder research platform was constructed; (b) exhaust gas speciation was performed to characterize the nature of hydrocarbon emissions from hydrogen-enriched natural gas fuels; (c) three H{sub 2}/CH{sub 4} fuel compositions were analyzed using spark timing and equivalence ratio sweeping procedures and finally; (d) a full size pick-up truck platform was converted to run on HCNG fuels. The testing performed in year one of the three year plan represents a baseline from which to assess options for overcoming the stated barriers to success.

Roser, R. [NRG Technologies, Inc., Reno, NV (United States)

1998-08-01

393

Reformer Fuel Injector  

Science.gov (United States)

Today's form of jet engine power comes from what is called a gas turbine engine. This engine is on average 14% efficient and emits great quantities of green house gas carbon dioxide and air pollutants, Le. nitrogen oxides and sulfur oxides. The alternate method being researched involves a reformer and a solid oxide fuel cell (SOFC). Reformers are becoming a popular area of research within the industry scale. NASA Glenn Research Center's approach is based on modifying the large aspects of industry reforming processes into a smaller jet fuel reformer. This process must not only be scaled down in size, but also decrease in weight and increase in efficiency. In comparison to today's method, the Jet A fuel reformer will be more efficient as well as reduce the amount of air pollutants discharged. The intent is to develop a 10kW process that can be used to satisfy the needs of commercial jet engines. Presently, commercial jets use Jet-A fuel, which is a kerosene based hydrocarbon fuel. Hydrocarbon fuels cannot be directly fed into a SOFC for the reason that the high temperature causes it to decompose into solid carbon and Hz. A reforming process converts fuel into hydrogen and supplies it to a fuel cell for power, as well as eliminating sulfur compounds. The SOFC produces electricity by converting H2 and CO2. The reformer contains a catalyst which is used to speed up the reaction rate and overall conversion. An outside company will perform a catalyst screening with our baseline Jet-A fuel to determine the most durable catalyst for this application. Our project team is focusing on the overall research of the reforming process. Eventually we will do a component evaluation on the different reformer designs and catalysts. The current status of the project is the completion of buildup in the test rig and check outs on all equipment and electronic signals to our data system. The objective is to test various reformer designs and catalysts in our test rig to determine the most efficient configuration to incorporate into the specific compact jet he1 reformer test rig. Additional information is included in the original extended abstract.

Suder, Jennifer L.

2004-01-01

394

Nuclear fuel handling apparatus  

International Nuclear Information System (INIS)

A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

395

Reactor fuel exchanging facility  

International Nuclear Information System (INIS)

Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

396

Failed fuel monitor  

International Nuclear Information System (INIS)

In a reactor in which the height of the upper end of a fuel assembly and that of the upper end of an upper lattice plate are identical, a sipping device capable of confirming that the water level is lower than the upper end of the fuel assembly is provided. That is, an existent sipper cap had a box-like shape, but it is modified to have such a shape that a portion of the sipper cap enters downward through the gap between the fuel assembly and the upper lattice plate. Then, an air escape hole is perforated at the entire portion. With such a constitution, if the water level is reduced to lower than the air escape hole, air comes out of the sipper cap thereby enabling to confirm the water level. Accordingly, even in a new type reactor in which the height of the upper end of the fuel assembly is identical with that of the upper end of the lattice plate, it is confirmed that other coolants are isolated by determining the water level to be lower than the upper end of the fuel assembly. (I.S.)

397

Fuel oil compositions  

Energy Technology Data Exchange (ETDEWEB)

An additive is provided for a fuel oil, especially middle distillate fuel oils and diesel fuel. The additive has been found to be effective in reducing sediment and gum formation during storage, and also reduces the coking of fuel injector nozzles when used in diesel engines. According to the invention, the fuel oil contains a minor proportion by weight of a mixture of 20-40 wt% of a polyphenol, sulfurized polyphenol, or a hindered phenol, and 80-60 wt% of a cyclic amide derived from a dicarboxylic acid or anhydride having a H- and C-containing substituent of at least 40 carbon atoms and a polyalkylene polyamine having at least 2 nitrogen atoms and at least 3 carbon atoms (other than carbon atoms in the branched substituents between the terminal acid groups). A suitable additive is 70 wt% macrocyclic derivative of polyisobutanyl succinic anhydride and penta propylene hexamine, and 30 wt% 4,4{sup 1} methylene bis (2,6 di tert butyl phenol). Experiments are described to illustrate the invention. 5 tabs.

Blackshaw, H.E.; Claydon, D.J.; Taylor, M.G.; Ilnycknj, S.

1990-06-26

398

Nuclear fuel pellets  

International Nuclear Information System (INIS)

Purpose: To suppress the generation of stress corrosions in the inner surface of fuel cans. Constitution: To a fuel pellet disk having a height smaller than the outer diameter, a coaxial V-shaped groove having a depth of about 1/4-1/3 of the pellet height is formed each in the upper end face and the lower end face, coaxially between the central region and the outer circumferential region and about at the radially center position. Further, a plurality of radial V-shaped grooves each having a depth of about 1/4-1/3 of the pellet height are formed both in the upper end face and the lower end face at the opposing positions and in the central region and the outer circumferential region while spaced apart radially by the coaxial V-shaped groove and displaced circumferentially. In the pellet of such a structure, if fuel-cladding interactions are caused during reactor operation, since the cracks formed in the pellets are radially disconnected at the boundary between the central region and the outer circumferential region, the heat radiation at high temperature, gas flow of fission products at high density and gamma rays issued from the center of the pellet are moderated till they arrive at the inner surface of the fuel can to thereby prevent the local stress corrosions in the inner surface of the fuel can. (Moriyama, K.)

399

Fuel from ashes  

International Nuclear Information System (INIS)

Since the middle of the fifties, chemistry and process technology are concerned with reprocessing nuclear fuels, i.e. with the chemical separation of radio-active waste materials of the spent fuel rods removed from a nuclear reactor. This optimization of the fuel cycle, today, is closely linked to the public discussion about the future nuclear energy policy in the Federal Republic of Germany: the issue of waste disposal with nuclear power stations has obtained a central position. The discussion of technical aspects has been expanded by a discussion of political and economic aspects: Does reprocessing make any sense at all, espec. with a view to economy. What will be the quantity of spent fuel elements obtaining and what dimension is purposeful for selecting plant capacity. Will they be manageable in terms of safety technology and will they be usable without unadmissible hazards and risks to man and environment. Where shall radioactive waste be disposed of. In this book, the authors give a representation of experience gained with reprocessing spent nuclear fuels: technical steps of reprocessing; operational experience with reprocessing plants; operation breakdowns and incidents; transgressing aspects like nuclear material surveillance, work protection, radio-ecology, ultimate storage; development chances of the PUREX process. (orig.)

400

Apparatus for fuel replacement  

International Nuclear Information System (INIS)

Object: To support a telescope mast such that no deforming load is applied to it even during massive vibration, it is held fixed at the time of fuel replacement to permit satisfactory remote control operation by automatic operation. Structure: The body of the fuel replacement apparatus is provided with telescope mast fixing means comprising a slide base supported for reciprocal movement with respect to a telescope mast, an operating arm pivoted at the slide base, a wrist member mounted on the free end of the operating arm and an engagement member for restricting the slide base and operating arm at the time of loading and unloading the fuel. When loading and unloading the fuel, the slide base and operating arm are restrained by the engagement member to reliably restrict the vibration of the telescope mast. When the fuel replacement apparatus is moved, the means provided on the operating arm is smoothly displaced to follow the swing (vibration) of the telescope mast to prevent the deforming load from being applied to the support portion or other areas. The wrist member supports the telescope mast such that it can be rotated while restraining movement in the axial direction, and it is provided with revolution drive means for rotating the telescope mast under remote control. (Kamimura, M.)

401

Fueling Global Fishing Fleets  

International Nuclear Information System (INIS)

Over the course of the 20th century, fossil fuels became the dominant energy input to most of the world's fisheries. Although various analyses have quantified fuel inputs to individual fisheries, to date, no attempt has been made to quantify the global scale and to map the distribution of fuel consumed by fisheries. By integrating data representing more than 250 fisheries from around the world with spatially resolved catch statistics for 2000, we calculate that globally, fisheries burned almost 50 billion L of fuel in the process of landing just over 80 million t of marine fish and invertebrates for an average rate of 620 L/t. Consequently, fisheries account for about 1.2% of global oil consumption, an amount equivalent to that burned by the Netherlands, the 18th-ranked oil consuming country globally, and directly emit more than 130 million t of CO2 into the atmosphere. From an efficiency perspective, the energy content of the fuel burned by global fisheries is 12.5 times greater than the edible protein energy content of the resulting catch

402

Liquid fuel from biomass  

International Nuclear Information System (INIS)

Various options for Danish production of liquid motor fuels from biomass have been studied in the context of the impact of EEC new common agricultural policy on prices and production quantities of crops, processes and production economy, restraints concerning present and future markets in Denmark, environmental aspects, in particular substitution of fossil fuels in the overall production and end-use, revenue loss required to assure competition with fossil fuels and national competence in business, industry and research. The options studied are rapeseed oil and derivates, ethanol, methanol and other thermo-chemical conversion products. The study shows that the combination of fuel production and co-generation of heat and electricity carried out with energy efficiency and utilization of surplus electricity is important for the economics under Danish conditions. Considering all aspects, ethanol production seems most favorable but in the long term, pyrolyses with catalytic cracking could be an interesting option. The cheapest source of biomass in Denmark is straw, where a considerable amount of the surplus could be used. Whole crop harvested wheat on land otherwise set aside to be fallow could also be an important source for ethanol production. Most of the options contribute favorably to reductions of fossil fuel consumption, but variations are large and the substitution factor is to a great extent dependent on the individual case. (AB) (32 refs.)

403

Metal fuel for reactor  

International Nuclear Information System (INIS)

The present invention concerns nuclear metal fuels suitable to FBR type reactors and the object of the invention is to prevent partial reduction of melting point due to the re-distribution of the ZR ingredient in the metal fuels comprising U-Pu-Zr system alloy or U-Zr system alloy. That is, at a radial intermediate portion in a radial cross section of a cylindrical metal fuel comprising U-Pu-Zr system alloy or U-Zr system, a region with higher Zr content than that in the U-Pu-Zr system alloy or U-Zr system alloy is disposed. In conventional metal fuels of this type, the content at the radial intermediate portion of the cylinder is reduced accompanying the advance of burning and re-distributed such that the content ratio between the central region and the outer circumferential region is increased. On the other hand, in the fuels having the feature of the present invention, the higher Zr content region constitutes a source for the Zr ingredient even if the re-distribution of the alloy ingredient occurs, thereby enabling to prevent the reduction of the Zr ingredient. (I.S.)

404

Bio-fuels  

International Nuclear Information System (INIS)

This report presents an overview of the technologies which are currently used or presently developed for the production of bio-fuels in Europe and more particularly in France. After a brief history of this production since the beginning of the 20. century, the authors describe the support to agriculture and the influence of the Common Agricultural Policy, outline the influence of the present context of struggle against the greenhouse effect, and present the European legislative context. Data on the bio-fuels consumption in the European Union in 2006 are discussed. An overview of the evolution of the activity related to bio-fuels in France, indicating the locations of ethanol and bio-diesel production facilities, and the evolution of bio-fuel consumption, is given. The German situation is briefly presented. Production of ethanol by fermentation, the manufacturing of ETBE, the bio-diesel production from vegetable oils are discussed. Second generation bio-fuels are then presented (cellulose enzymatic processing), together with studies on thermochemical processes and available biomass resources

405

Safety analysis of MOX fuels by fuel performance code  

Energy Technology Data Exchange (ETDEWEB)

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

406

Safety analysis of MOX fuels by fuel performance code  

International Nuclear Information System (INIS)

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

407

Low contaminant formic acid fuel for direct liquid fuel cell  

Science.gov (United States)

A low contaminant formic acid fuel is especially suited toward use in a direct organic liquid fuel cell. A fuel of the invention provides high power output that is maintained for a substantial time and the fuel is substantially non-flammable. Specific contaminants and contaminant levels have been identified as being deleterious to the performance of a formic acid fuel in a fuel cell, and embodiments of the invention provide low contaminant fuels that have improved performance compared to known commercial bulk grade and commercial purified grade formic acid fuels. Preferred embodiment fuels (and fuel cells containing such fuels) including low levels of a combination of key contaminants, including acetic acid, methyl formate, and methanol.

Masel, Richard I. (Champaign, IL); Zhu, Yimin (Urbana, IL); Kahn, Zakia (Palatine, IL); Man, Malcolm (Vancouver, CA)

2009-11-17

408

Fuel development program of the nuclear fuel element centre  

International Nuclear Information System (INIS)

Fuel technology development program pf the nuclear fuel element centre is still devised into two main pillars, namely the research reactors fuel technology and the power reactor fuel technology taking into account the strategic influencing environment such as better access to global market of fuel cycle services, the state of the art and the general trend of the fuel technology in the world. Embarking on the twenty first century the fuel development program has to be directed toward strengthening measure to acquire and self-reliance in the field of fuel technology in support to the national energy program as well as to the utilisation of research reactor. A more strengthened acquisition of fuel cycle technology, in general, and particularly of fuel technology would improve the bargaining power when negotiation the commercial fuel technology transfer in the future

409

Fuel-cycle cost comparisons with oxide and silicide fuels  

International Nuclear Information System (INIS)

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

410

FUEL CELLS IN ENERGY PRODUCTION  

OpenAIRE

The purpose of this thesis is to study fuel cells. They convert chemical energy directly into electrical energy with high efficiency and low emmission of pollutants. This thesis provides an overview of fuel cell technology.The basic working principle of fuel cells and the basic fuel cell system components are introduced in this thesis. The properties, advantages, disadvantages and applications of six different kinds of fuel cells are introduced. Then the efficiency of each fuel cell is p...

Huang, Xiaoyu

2011-01-01

411

COSIMA fuel rod simulator  

International Nuclear Information System (INIS)

This paper presents a description of the fuel rod simulator with an indirect electrical heating used in the COSIMA blowdown test facility at the Kernforschungszentrum Karlsruhe (KfK). Experiments with fuel rod simulators are being conducted in the COSIMA test facility to investigate cladding tube behavior during LOCA (Loss-of-coolant accident) conditions. Cladding tubes identical to those of a PWR (Pressurized Water Reactor) are used in the simulator design. The heater element is a graphite rod located in the center of the simulator. The thermal behavior of the nuclear fuel is simulated by annular alumina pellets or thoria pellets. The maximum rod power is 730 W/cm. To date, about 100 blowdown transients with 35 single rods have been run. The cladding surface temperature is measured by pyrometers via light guides. The design of the rods, the operating behavior, the failure analysis, and the pellet behavior are described

412

Fuel cells flows study  

International Nuclear Information System (INIS)

Fuel cells are energy converters, which directly and continuously produce electricity from paired oxidation reduction-reactions: In most cases, the reactants are oxygen and hydrogen with water as residue. There are several types of fuel cells using various electrolytes and working at different temperatures. Proton Exchange Membrane Fuel Cells are, in particular, studied in the GESTEAU facility. PEMFC performance is chiefly limited by two thermal-hydraulic phenomena: the drying of membranes and the flooding of gas distributors. Up to now, work has been focused on water flooding of gas channels. This has showed the influence of flow type on the electrical behaviour of the cells and the results obtained have led to proposals for new duct geometries. (authors)

413

Fuel cells in transportation  

Energy Technology Data Exchange (ETDEWEB)

A promising new power source for electric drive systems is the fuel cell technology with hydrogen as energy input. The worldwide fuel cell development concentrates on basic research efforts aiming at improving this new technology and at developing applications that might reach market maturity in the very near future. Due to the progress achieved, the interest is now steadily turning to the development of overall systems such as demonstration plants for different purposes: electricity generation, drive systems for road vehicles, ships and railroads. This paper does not present results concerning the market potential of fuel cells in transportation but rather addresses some questions and reflections that are subject to further research of both engineers and economists. Some joint effort of this research will be conducted under the umbrella of the IEA Implementing Agreement 026 - Annex X, but there is a lot more to be done in this challenging but also promising fields. (EG) 18 refs.

Erdmann, G. [Technische Univ., Berlin (Germany); Hoehlein, B. [Research Center Juelich (Germany)

1996-12-01

414

HTGR fuel cycle  

International Nuclear Information System (INIS)

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

415

Fuel cell. Nenryo denchi  

Energy Technology Data Exchange (ETDEWEB)

In the present invention, there are provided plural number of unit cells having an electrolyte between a fuel electrode and an oxidant electrode, a gas separator plate having gas-supply grooves insertingly placed between the unit cells, and metal-sheathed thermocouple thermometer which are located n said gas-supply grooves and are applied with afluorine resin coating. By this arrangement, a wire-breakage at the temperature measurement is prevented giving a fuel cell with high reliability. In the conventional fuel cell, a thermocouple thermometer is only a fluorine resin-coated naked thermocouple and is easily broken because of lack of strength. This invention with a metal-sheathed thermocouple placed in a gas-supply groove considerably reduced the danger of wire breakage. 4 figs.

Shinohara, Y.; Nishizawa, N. (Mitsubishi Electric Engineering Co. Ltd., Tokyo (Japan)); Taniguchi, T. (Mitsubishi Electric Corp., Tokyo (Japan))

1991-05-22

416

Nuclear fuel strategies  

International Nuclear Information System (INIS)

The paper reports on two international meetings on nuclear fuel strategies, one organised by the World Nuclear Fuel Market in Seville (Spain) October 1988, and the other organised by the American and European nuclear societies in Washington (U.S.A.) November 1988. At the Washington meeting a description was given of the uranium supply and demand market, whereas free trade in uranium was considered in Seville. Considerable concern was expressed at both meetings on the effect on the uranium and enrichment services market of very low prices for spot deals being offered by China and the Soviet Union. Excess enrichment capacity, the procurement policies of the USA and other countries, and fuel cycle strategies, were also discussed. (U.K.)

417

Automated breeder fuel fabrication  

International Nuclear Information System (INIS)

The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

418

KUR spent fuel handling  

International Nuclear Information System (INIS)

The spent fuel elements of HEU (253 as of July 1997 and 389 as of April 2004) will be sent back to US by March 2006. In July 1997, the contract (No. DE-AC09-97SR18907) was completed between DOE and Kyoto University. One or two casks will be made in 1998 and the shipment will start in 1999. So many paper works and negotiations with the Government of Japan and the local governments have been aggressively executed by committee members of Spent Fuel Countermeasure. One of the highlights is to find a port of shipment. Technical and political problems including public acceptance are to be solved. In this paper, mainly technical matters related to spent fuel handling are described. (author)

419

Fuel Element Technical Manual  

Energy Technology Data Exchange (ETDEWEB)

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

Burley, H.H. [ed.

1956-08-01

420

Nuclear fuel assembly  

International Nuclear Information System (INIS)

A filter structure having inflow holes or outflow holes of a coolant channel is disposed, in the direction perpendicular to the flowing direction of coolants, to the coolant flow channel for allowing coolants to pass through in a nuclear fuel rod support lattice of a lower tie plate of a nuclear fuel assembly. Ascending coolants flow into the inflow holes or outflow holes on a side wall and obstacles pass a flow line different from the flow line of the coolants and can not intrude but accumulated in a lower tie plate. In addition, since coolants inflow from all directions into cylinders of the filter structure, obstacles are less accumulated to the inflow holes or outflow holes. Accordingly, such clogging of the flow channel that obstacles close the inflow holes is not caused. Intrusion of obstacles is prevented, and failure or fretting abrasion of nuclear fuel rods or spacers is prevented. (N.H.)

421

Nuclear fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly for a BWR type reactor, a groove is disposed to the lower end plug of the fuel assembly to form a cooling water channel between the lower end plug and a lower plate. As a result, temperature elevation of the lower end plug due to heat conduction from a fuel portion can be suppressed. The corrosion on the surface of the lower end plug caused by a long use in the reactor can be suppressed and the adhesion sticking between the end plug insertion hole and the lower end plug can be prevented. Further, the contact area is remarkably reduced by disposing the groove, and no great force such as applying damages upon withdrawal is not required even if they should be sticked. (N.H.)

422

Casket for reactor fuel  

International Nuclear Information System (INIS)

Casket for reactor fuel provided with many vertically arranged bars of fuel, a small number of which being absorption bars containing a poison in the form of Gd2O3 and other in greater number which are free of poison, each absorption bar containing principally a number of combustible bodies arranged axially directly one behind the other within a capsule tube, some of them being free of poison while others contain poison, each absorption bar offering for each absorption zone in the fuel casket a number of successive axially arranged sections of zone and the average amount of poison per volume unit varying between the different sections, characterized in that as regards the amount of poison per volume unit the number of sections in an absorption bar is greater that the number of the different combustible bodies in the bar

423

Weight simulation fuel assembly  

International Nuclear Information System (INIS)

A tungsten pellet is not applied with hollow fabrication but a tungsten rod is inserted and filled into a zircaloy fuel cladding tube, as well as different kind of material having a density lower than that of tungsten, for example, stainless steel rods, are disposed successively intermittently and alternately for simulating the weight of one fuel rod. The filling method and the length of the individual pellets are optional depending on the method of usage, providing that the outer diameter of the simulation pellet is made identical with that of the actual fuel pellet. With such a constitution, there is no need to dispose a hollow portion as in the case of using only tungsten pellets, and the costs for both the materials and the fabrication can be saved. (T.M.)

424

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To provide fuel rods for use in BWR type reactors with such structures as having satisfactory gap conductance even at a high burning degree. Constitution: Hollow pellets of uranium dioxide and plutonium dioxide are charged in a cylindrical metal cladding tube and metal rods containing (n, ?) reaction materials are inserted into the hollow portion to constitute a fuel rod. Iron or steel containing, for example, 0.3 % boron 10 is suitable as the metal rod. As a result, helium required for preventing the reduction in the gap conductance due to the release of gaseous fission products can be supplied even at high burning degree. Since no local changes are resulted to the neutron distribution or power distribution and thermal neutron absorption occurs at the center of the pellets, the radial temperature distribution in the fuel pellets can be flattened, and the axial temperature distribution is also flattened by the highly heat conductive metal rods. (Horiuchi, T.)

425

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To provide hollow pellet type nuclear fuel elements each being provided with a mechanism for preventing the bouncing of pellets and the falling of broken pieces of pellets into central holes and, simultaneously, even when a PCI occurs, excessive stresses are not generated in the cladding tube. Constitution: Each of nuclear fuel elements is composed of a cladding tube both ends of which are tightly closed, and a plurality of hollow pellets charged in the cladding tube. In the fuel elements constituted as described above, there are laminated cylindrical bodies liable to be deformed more easily than the cladding tube in the holes of said hollow pellets, thereby preventing the bouncing of the pellets. (Aizawa, K.)

426

Fuel-coolant interactions  

International Nuclear Information System (INIS)

An important aspect of nuclear fuel behaviour that impacts on the fuel cycle is the interaction of the cladding with the coolant. In particular, the accumulation of deposited crud (corrosion products transported in the reactor coolant) on fuel element surfaces can severely hamper fuel performance by impeding heat transfer and promoting cladding corrosion, both of which may lead to fuel defects and the release of fission products and actinides to the primary coolant systems. Crud deposition is therefore an important consideration in reactor operation; it not only leads to poor performance and radiation field growth by exacerbating fuel defects but also serves as the source of radionuclides such as Co-60 which are major contaminants of out-reactor components. Furthermore, the sequestering of boron from the coolant by fuel deposits in PWRs can give rise to control problems as reactor flux characteristics are modified. As utilities apply the ALARA principle (As Low As Reasonably) to the management of occupational radiation doses and at the same time endeavour to optimise the fuel cycle, it becomes clear that an understanding of the mechanisms involved in coolant-cladding interactions is vital. There are several mechanisms of interest here. The source of crud is the fundamental corrosion process accruing on surfaces of the coolant system and the interaction of that process with local regimes of coolant flow. Accordingly, differences in the chemical and physical condition of the coolant across the reactor core and the steam generators are important factors in CANDUs and PWRs determining release of corrosion products from surfaces, while similar processes along the feedtrain influence crud levels in the reactor coolant in BWRs. The nature of suspended crud, which is determined by the materials of construction of the various components of the coolant system and the chemistry control of the coolant itself, determines the interaction with fuel cladding. Thus, crud in CANDUs is dominated by iron oxide (magnetite Fe3O4) because of the large proportion of carbon steel in the circuit, while in PWRs crud is an iron-nickel oxide (nickel ferrite - NiFe2O4 or a variant) because of the presence of stainless steel and nickel alloys. The more oxidizing nature of BWR coolant causes a higher phase of iron oxide to occur, so that haematite (Fe2O3) becomes a constituent of deposits in BWRs. The deposition of the suspended crud of fuel surfaces is influenced by the electrostatic charge on the particles themselves and on the fuel surface. The chemistry regime-oxidizing nature, alkalinity, boron concentration, etc.-determines those surface charges. The forces arising from the thermalhydraulic conditions in the core and the physical properties of the crud (such as particle size) then interact with the surface forces to determine the deposition characteristics. Besides the deposition of suspended material, the deposition of corrosion products from solution can occur and in fact may dominate in CANDUs and PWRs where the solubility of oxides relatively high. In that case, it is important to tailor the coolant chemistry to minimize the solubility and to ensure that the change of solubility with temperature is not such as to promote massive precipitation in the core. Even then, adsorption-desorption at fuel surfaces of ions such as Co2+ will lead to a level of system activation that depends on the indigenous corrosion film on the cladding surface

427

Fuel cells I  

Energy Technology Data Exchange (ETDEWEB)

This book contains six contributions in the field of proton-conducting membranes for fuel cells: 1. A proton-conducting polymer membrane as solid electrolyte - Function and required properties (Lorenz Gubler and Guenther G. Scherer); 2. Proton-conducting polymer electrolyte membranes: Water and structure in charge (Michael Eikerling, Alexei A. Kornyshev and Eckhard Spohr); 3. structural and morphological features of acid-bearing polymers for PEM fuel cells (Yunsong Yang, Ana Siu, Timothy J. Peckham and Steven Holdcroft); 4. Perfluorinated ionic polymers for PEFCs (including supported PFSA) (M. Yoshitake and A. Watakabe); 5. Radiation grafted membranes (Selmiye Alkan Guersel, Lorenz Gubler, Bhuvanesh Gupta and Guenther G. Scherer); 6. Advances in the development of inorganic-organic membranes for fuel cell applications (Deborah J. Jones and Jacques Roziere).

Scherer, Guenther G. (ed.) [Paul Scherrer Institut, Villigen (Switzerland). Labor fuer Elektrochemie

2008-07-01

428

Compact fuel cell  

Science.gov (United States)

A novel electrochemical cell which may be a solid oxide fuel cell (SOFC) is disclosed where the cathodes (144, 140) may be exposed to the air and open to the ambient atmosphere without further housing. Current collector (145) extends through a first cathode on one side of a unit and over the unit through the cathode on the other side of the unit and is in electrical contact via lead (146) with housing unit (122 and 124). Electrical insulator (170) prevents electrical contact between two units. Fuel inlet manifold (134) allows fuel to communicate with internal space (138) between the anodes (154 and 156). Electrically insulating members (164 and 166) prevent the current collector from being in electrical contact with the anode.

Jacobson, Craig (Moraga, CA); DeJonghe, Lutgard C. (Lafayette, CA); Lu, Chun (Richland, WA)

2010-10-19

429

Fuel cell membrane humidification  

Science.gov (United States)

A polymer electrolyte membrane fuel cell assembly has an anode side and a cathode side separated by the membrane and generating electrical current by electrochemical reactions between a fuel gas and an oxidant. The anode side comprises a hydrophobic gas diffusion backing contacting one side of the membrane and having hydrophilic areas therein for providing liquid water directly to the one side of the membrane through the hydrophilic areas of the gas diffusion backing. In a preferred embodiment, the hydrophilic areas of the gas diffusion backing are formed by sewing a hydrophilic thread through the backing. Liquid water is distributed over the gas diffusion backing in distribution channels that are separate from the fuel distribution channels.

Wilson, Mahlon S. (Los Alamos, NM)

1999-01-01

430

Fuel cycle based safeguards  

International Nuclear Information System (INIS)

In NPT safeguards the same model approach and absolute-quantity inspection goals are applied at present to all similar facilities, irrespective of the State's fuel cycle. There is a continuing interest and activity on the part of the IAEA in new NPT safeguards approaches that more directly address a State's nuclear activities as a whole. This fuel cycle based safeguards system is expected to a) provide a statement of findings for the entire State rather than only for individual facilities; b) allocate inspection efforts so as to reflect more realistically the different categories of nuclear materials in the different parts of the fuel cycle and c) provide more timely and better coordinated information on the inputs, outputs and inventories of nuclear materials in a State. (orig./RF)

431

Nuclear fuel element  

International Nuclear Information System (INIS)

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

432

Spent fuel pyroprocessing demonstration  

International Nuclear Information System (INIS)

A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

433

AMONCO. Biogas fuel cells  

Energy Technology Data Exchange (ETDEWEB)

The usage of biogas in fuel cells (FC) is the primary goal of the AMONCO project. The precondition for the use of biogas in FCs is the avoidance, or elimination respective reduction of detrimental trace gases, which are potentially harmful for fuel cells. Referring to that AMONCO project has the following core objectives: 1) Comprehensive biogas analyses in quality and quantity on a detailed level - identifying harmful trace gases for fuel cells, 2) avoidance of detrimental trace gases in biogas through optimal composition of the feedstock, 3) advanced controlling of the anaerobic digestion process to hinder the formation of trace gases while keeping a high CH{sub 4} yield, 4) suitable and cost-effective biogas cleaning towards the utilisation in fuel cells, 5) investigation and assessment of the effects of biogas in fuel cells through single cell tests and 6) development of techno- and socio-economic 'market driven' implementation strategies. The major work tasks to fulfill these objectives include the development of a knowledge based decision support tool with the capability to predict trace gases in dependence of the fermented substrates and a cost-effective cleaning process removing the significant trace gases detrimental for fuel cell systems. The decision support tool assists on the one hand the operators of biogas plants in the selection of composition of input substrate causing the lowest possible concentration of trace gases. On the other hand the decision support tool provides the ability of the in-situ control of anaerobic digestion (AD) process towards lowest concentration of trace gases while keeping a maximum yield of CH{sub 4}. (O.M.)

Simader, G.R.

2003-09-01

434

Direct Methanol Fuel Cell, DMFC  

Directory of Open Access Journals (Sweden)

Full Text Available Direct Methanol Fuel Cell, DMFC is a kind of fuel cell using methanol as a fuel for electric producing. Methanol is low cost chemical substance and it is less harmful than that of hydrogen fuel. From these reasons it can be commercial product. The electrocatalytic reaction of methanol fuel uses Pt-Ru metals as the most efficient catalyst. In addition, the property of membrane and system designation are also effect to the fuel cell efficient. Because of low power of methanol fuel cell therefore, direct methanol fuel cell is proper to use for the energy source of small electrical devices and vehicles etc.

Amornpitoksuk, P.

2003-09-01

435

Alternate fusion fuels workshop  

Energy Technology Data Exchange (ETDEWEB)

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

1981-06-01

436

Linseed as renewable fuel  

International Nuclear Information System (INIS)

The cultivation of renewable fuels is a way of reducing excess food production and of opening up alternatives of labour and income on the agricultural sector. Industry takes an interest in renewable fuels because of the ecological aspect. Vegetable oils may be used as lubricants (e.g. for chainsaws, hydraulic systems and two-stroke engines), while starck may be utilized as biodegradable packaging material. The report therefore investigates the chances for linseed production in Germany. The economic efficiency of linseed production may be improved by utilizing the slow-degradable linseed stran as mulding material in gardening and landscaping and for erosion protection. (orig.)

437

Nuclear fuel cladding tubes  

International Nuclear Information System (INIS)

Purpose: To maintain the poisoning effect for a required period of time and moderate the stresses due to the fuel-cladding interactions in cladding tubes for gadolinium oxide-containing UO2 fuels. Constitution: The cladding tube is made of a zirconium alloy. A metal or alloy compatible with and softer than the zirconium alloy is lined to the inner surface of the tube. Further, the lined layer is incorporated with burnable poisons selected from the group consisting of Cd, Sm, Dy, Du and B. (Ikeda, J.)

438

Fuels from renewable resources  

Science.gov (United States)

Consideration is given to fuel substitution based on regenerative plants. Methanol can be produced from regenerative plants by gasification followed by the catalytic hydration of carbon oxides. Ethanol can be used as a replacement fuel in gasoline and diesel engines and its high-knock rating allows it to be mixed with lead-free gasoline. Due to the depletion of oil and gas reserves, fermentation alcohol is being considered. The raw materials for the fermentation process can potentially include: (1) sugar (such as yeasts, beet or cane sugar); (2) starch (from potatoes or grain) and (3) cellulose which can be hydrolized into glucose for fermentation.

Hoffmann, L.; Schnell, C.; Gieseler, G.

439

Nuclear reactor fuel rod  

International Nuclear Information System (INIS)

The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG)

440

Transporting uranium fuels  

International Nuclear Information System (INIS)

The transport of nuclear materials is often overlooked when viewing the front-end of the nuclear fuel cycle. However, when book transfers are not practical, a well-managed transport program is a significant factor in customer satisfaction for nuclear fuel purchases. Some of the logistical considerations of transporting uranium concentrates and uranium hexafluoride include which type of cylinders to use for transport, modes of transport available, sampling, shipment scheduling, and coordination of transportation services. Further consideration must be given to the numerous domestic and foreign regulations, fees, and issues of insurance; however, the complexity of these latter issues prevents their complete analysis in this discussion

441

Bioethanol: fuel or feedstock?  

DEFF Research Database (Denmark)

Increasing amounts of bioethanol are being produced from fermentation of biomass, mainly to counteract the continuing depletion of fossil resources and the consequential escalation of oil prices. Today, bioethanol is mainly utilized as a fuel or fuel additive in motor vehicles, but it could also be used as a versatile feedstock in the chemical industry. Currently the production of carbon-containing commodity chemicals is dependent on fossil resources, and more than 95% of these chemicals are produced from non-renewable carbon resources. The question is: what will be the optimal use of bioethanol in a longer perspective? (c) 2007 Society of Chemical Industry.

Rass-Hansen, Jeppe; Falsig, Hanne

2007-01-01

442

Clean fuels from biomass  

Science.gov (United States)

The paper discusses the U.S. resources to provide fuels from agricultural products, the present status of conversion technology of clean fuels from biomass, and a system study directed to determine the energy budget, and environmental and socioeconomic impacts. Conversion processes are discussed relative to pyrolysis and anaerobic fermentation. Pyrolysis breaks the cellulose molecules to smaller molecules under high temperature in the absence of oxygen, wheras anaerobic fermentation is used to convert biomass to methane by means of bacteria. Cost optimization and energy utilization are also discussed.

Hsu, Y.-Y.

1976-01-01

443

Synthetic fuel composition  

Energy Technology Data Exchange (ETDEWEB)

A composition useful as a synthetic fuel for fireplaces and the like is described. The composition contains particulate coal as the major component in combination with slack wax, sodium and/or potassium silicate and an oxidizing agent. Minor amounts of coloring agents or agents for providing desired aromas or the like are optionally included in the composition. The composition is conveniently provided in the form of a log which can be used as a synthetic fuel. The composition burns slowly and evenly without excessive smoking while substantially retaining its original shape thereby facilitating removal of the resultant ashes.

Anderson, T.J.

1981-04-07

444

AAA fuels handbook  

International Nuclear Information System (INIS)

PART A of this handbook is for metal alloy fuels. The metal alloy of transuranic elements (i.e., Pu, Np, Am, Cm, etc) in Zr, designated as TRU-Zr, is one of the primary candidate fuel types for the AAA system. The data found in the literature were critically reviewed and assessed to provide the recommended ones. For the convenience of the user, most of the materials properties are given in model correlations; performance models are also provided in mathematical formulas. Tabulations were made in case where these were judged to allow more flexibility for the user. The information for the materials properties of the TRU-Zr alloy, however, is extremely scarce in general. Therefore, where no data exists, the values and models based on theoretical estimations and extrapolations from the U-Zr and U-Pu-Zr data are inevitably recommended. The justifications for this will be possible when sufficient measured data are available in the future. In this respect, this part is subject to modification whenever new data or better methods of deduction become available. The purpose of PART B is to provide the best available fuel materials properties and performance models of the (Pu,Zr)N and (TRU,Zr)N solid-solution fuels for fuel design and safety calculation of the Advanced Accelerator Assisted (AAA) system. PART B parallels PART A, Metal Alloy Fuels, in form and topics. The solid solution of the mononitrides of transuranic elements (i.e., Pu, Np, Am, Cm, etc) with zirconium mononitri Np, Am, Cm, etc) with zirconium mononitride, designated as (TRU,Zr)N, is one of the primary candidate fuel types for the AAA system, together with metallic TRU-Zr alloy fuels studied in Argonne National Laboratory. The data found in the literature were critically reviewed and assessed to provide the recommended ones. For the convenience of the user, most of the materials properties are given in model correlations; performance models are also provided in mathematical formulas. Tabulations were made in some cases where these were regarded to allow more flexibility for the user. The information for the materials properties of (Pu,Zr)N and (TRU,Zr)N, however, is in many cases nonexistent in general. Therefore, where no data exists, the values and models based on theoretical estimations are made. Extrapolations from the similar ceramic materials such as (U,Pu)C and (U,Pu)N, which have relatively more data, are inevitably relied upon. The justifications for this will be possible only when sufficient measured data become available in the future

445

75 FR 26049 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program  

Science.gov (United States)

...process heat fuel include biomass, biogas, coal, and natural gas. The information...for renewable fuel producers using biogas as process heat fuel would help EPA verify the contractual pathway of the biogas from the supplier to the...

2010-05-10

446

Study of transformations by annealing of the body. Centred cubic {gamma} phase of uranium-molybdenum alloys; Etude des transformations par revenu de la phase {gamma} cubique centree des alliages uranium-molybdene  

Energy Technology Data Exchange (ETDEWEB)

By annealing at different temperatures, we have studied the transformations of the body centred cubic {gamma} phase for two alloys containing 6 and 10 per cent molybdenum by weight respectively. There is a return to the equilibrium state by formation of the stable {alpha} orthorhombic and {epsilon} ordered tetragonal phases, following two types of reaction: - pearlite transformation by nucleation and growth from the grain boundaries, preponderant when the annealing takes place at temperature above 400 deg. C, and identical for the two types of alloys. This reaction has already been studied by numerous authors, who have constructed the corresponding TTT curves, - transformation inside the grains of the quenched solid solution when annealing takes place at 400 deg. C or below: 6 per cent alloy - precipitation of fine a phase particles, followed by progressive ordering of the solid solution enriched in molybdenum, 10 per cent alloy - formation of small ordered regions and then a fine a phase precipitate. In the course of this work we have paid particular attention to the study of intragranular reactions after low-temperature annealing, the reactions involved in this case not having been explained up to the present. The {gamma} phase transformation has been studied by means of three techniques: micrography - microhardness tests - X-ray diffraction. (author) [French] Nous avons etudie les transformations par revenu a differentes temperatures, de la phase {gamma} cubique centree des alliages U-Mo trempes, pour deux alliages a 6 et a 10 pour cent de molybdene en poids. Il y a retour a l'etat d'equilibre par formation des phases stables {alpha} orthorhombique et quadratique ordonnee, suivant deux types de reactions: - transformation perlitique par germination et croissance a partir des joints de grains, preponderante lorsque le recuit a lieu a temperature superieure a 400 deg. C, et identique pour les deux types d'alliages. Cette reaction a deja ete etudiee par de nombreux auteurs qui ont trace les courbes TTT correspondantes, - transformation a l'interieur des grains de la solution solide trempee lorsque le revenu a lieu a temperature inferieure ou egale a 400 deg. C: Alliage a 6 pour cent - precipitation de fines particules de phase {alpha}, puis mise en ordre progressive de la solution solide enrichie en molybdene. Alliage a 10 pour cent - formation de petits domaines ordonnes et ensuite d'un fin precipite de phase {alpha}. Au cours de ce travail, nous nous sommes specialement attaches a etudier les transformations intragranulaires apres recuit a basse temperature, les reactions intervenant dans ce cas n'ayant pas ete expliquees jusqu'ici. Nous avons etudie la transformation de la phase {gamma} au moyen des trois techniques: micrographie - microdurete - diffraction de rayons X. (auteur)

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

1959-06-15

447

Design package for fuel retrieval system fuel handling tool modification  

International Nuclear Information System (INIS)

This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

448

Heating subsurface formations by oxidizing fuel on a fuel carrier  

Science.gov (United States)

A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

Costello, Michael; Vinegar, Harold J.

2012-10-02

449

Design Package for Fuel Retrieval System Fuel Handling Tool Modification  

Energy Technology Data Exchange (ETDEWEB)

This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

TEDESCHI, D.J.

2000-03-27

450

Design Package for Fuel Retrieval System Fuel Handling Tool Modification  

International Nuclear Information System (INIS)

This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

451