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

As-cast uranium-molybdenum based metallic fuel candidates and the effects of carbon addition  

Science.gov (United States)

The objective of this research was to develop and recommend a metallic nuclear fuel candidate that lowered the onset temperature of gamma phase formation comparable or better than the uranium-10 wt. pct. molybdenum alloy, offered a solidus temperature as high or higher than uranium-10 wt. pct. zirconium (1250°C), and stabilized the fuel phase against interaction with iron and steel at least as much as uranium-10 wt. pct. zirconium stabilized the fuel phase. Two new as-cast alloy compositions were characterized to assess thermal equilibrium boundaries of the gamma phase field and the effect of carbon addition up to 0.22 wt. pct. The first system investigated was uranium- x wt. pct. M where x ranged between 5-20 wt. pct. M was held at a constant ratio of 50 wt. pct. molybdenum, 43 wt. pct. titanium, and 7 wt. pct. zirconium. The second system investigated was the uranium-molybdenum-tungsten system in the range 90 wt. pct. uranium - 10 wt. pct. molybdenum - 0 wt. pct. tungsten to 80 wt. pct. uranium - 10 wt. pct. molybdenum - 10 wt. pct. tungsten. The results showed that the solidus temperature increased with increased addition of M up to 12.5 wt. pct. for the uranium-M system. Alloy additions of titanium and zirconium were removed from uranium-molybdenum solid solution by carbide formation and segregation. The uranium-molybdenum-tungsten system solidus temperature increased to 1218°C at 2.5 wt. pct. with no significant change in temperature up to 5 wt. pct. tungsten suggesting the solubility limit of tungsten had been reached. Carbides were observed with surrounding areas enriched in both molybdenum and tungsten. The peak solidus temperatures for the alloy systems were roughly the same at 1226°C for the uranium-M system and 1218°C for the uranium-molybdenum-tungsten system. The uranium-molybdenum-tungsten system required less alloy addition to achieve similar solidus temperatures as the uranium-M system.

Blackwood, Van Stephen

3

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)

4

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

International Nuclear Information System (INIS)

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 monolithns. 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

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

Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels  

International Nuclear Information System (INIS)

Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the ? structure, - cooling rate at the transformation points, - whether or not the intermediate ? ? ? transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram ? + ?; ? + ? the effects of the morphology (in particular the two types of ? pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the ? structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors)

7

In-pile tests and post-reactor examinations of fuel elements with uranium-molybdenum fuel of reduced enrichment  

International Nuclear Information System (INIS)

The results of the in-pile tests and post-reactor examinations of dispersed fuel elements with fuel made of alloy U-9%Mo with 36% enrichment by 235U and uranium concentration of 5.4 g/cm3 manufactured by extrusion technique have been presented. The fuel elements have been tested during 107 effective days at IVV-2M reactor to reach average burnup of 40%. The maximum heat flux density was 0.69 MW/m2, while the maximum design temperature of fuel cladding has not exceeded 80 deg. C. The post-reactor examinations have been conducted using such techniques as visual inspection, profilometry, volume measurements, gamma spectroscopy, metallography and X-ray structure analysis. The following aspects have been examined: state of fuel elements; changes in shape, size and volume of fuel elements; structure and phase composition of the oxide film; character and depth of corrosion damage to fuel cladding; composition and structure of the dispersed fuel core; phase composition and width of the interaction areas of uranium-molybdenum fuel and the aluminium matrix. (author)

8

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)

9

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

International Nuclear Information System (INIS)

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

10

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

11

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

International Nuclear Information System (INIS)

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)

12

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

13

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining  

Digital Repository Infrastructure Vision for European Research (DRIVER)

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Fea...

Kleeck, Melissa A.

2011-01-01

14

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining  

Science.gov (United States)

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

Van Kleeck, Melissa A.

15

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)

16

Vitrification of HLW produced by uranium/molybdenum fuel reprocessing in cogema's cold crucible melter  

International Nuclear Information System (INIS)

The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficientlre glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs

17

Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter  

International Nuclear Information System (INIS)

The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficientlre glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs

18

Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo  

Energy Technology Data Exchange (ETDEWEB)

The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

Almeida, Cirila Tacconi de

2005-07-01

19

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor  

International Nuclear Information System (INIS)

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the developmenrovide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the bounds of known technology and are adaptable to the high-volume production required to process ? 2.5 to 4 tons of U/Mo and produce ?16,000 flat plates for U.S. reactors annually (?10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M

20

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor  

Energy Technology Data Exchange (ETDEWEB)

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the bounds of known technology and are adaptable to the high-volume production required to process {approx} 2.5 to 4 tons of U/Mo and produce {approx}16,000 flat plates for U.S. reactors annually ({approx}10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M.

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-09-30

 
 
 
 
21

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel  

Science.gov (United States)

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

22

Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter  

Energy Technology Data Exchange (ETDEWEB)

The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs.

Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod' homme, A.

2003-02-25

23

Set up of Uranium-Molybdenum powder production (HMD process)  

International Nuclear Information System (INIS)

Powder metallurgy offers different alternatives for the production of Uranium-Molybdenum (UMo) alloy powder in sizes smaller than 150 microns. This powder is intended to be used as a dispersion fuel in an aluminum matrix for research, testing and radioisotopes production reactors (MTR). A particular process of massive hydriding the UMo alloy in gamma phase has been developed. This work describes the final adjustments of process variables to obtain UMo powder by hydriding-milling-de hydriding (HMD) and its capability for industrial scaling up. (author)

24

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

25

Qualification of Uranium-Molybdenum Alloys for Research Reactor Community  

International Nuclear Information System (INIS)

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 effectives 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)

26

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

27

Production of uranium-molybdenum particles by spark-erosion  

International Nuclear Information System (INIS)

With the spark-erosion method we have produced spheroidal particles of an uranium-molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO2 with a distinctive size distribution with peaks centered at 70 and 10 ?m were obtained. The particles have central inclusions of U and Mo compounds

28

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)

29

Basic design of a rotating disk centrifugal atomizer for uranium-molybdenum alloys  

International Nuclear Information System (INIS)

One of the most used techniques to produce metallic powders is the centrifugal atomization with a rotating disk. This process is employ to fabricate ductile metallic particles of uranium-molybdenum alloys (typically U- 7 % Mo, by weight) for nuclear fuel elements for research and testing reactors. These alloys exhibit a face-centered cubic structure (? phase) which is stable above 700 C degrees and can be retained at room temperature. The rotating disk centrifugal atomization allows a rapid solidification of spherical metallic droplets of about 40 to 100 ?m, considered adequate to manufacture nuclear fuel elements. Besides the thermo-physical properties of both the alloy and the cooling gas, the main parameters of the process are the radius of the disk (R), the diameter of the atomization chamber (D), the disk rotation speed (?), the liquid volume flow rate (Q) and the superheating of the liquid (?T). In this work, they were applied approximate analytical models to estimate the optimal geometrical and operative parameters to obtain spherical metallic powder of U- 7 % Mo alloy. Three physical phenomena were considerate: the liquid metal flow along the surface of the disk, the fragmentation and spheroidization of the droplets and the cooling and solidification of the droplets. The principal results are the more suitable gas is helium; R ? 20 mm; D ? 1 m; ? 20,000 - 50,000 rpm; Q ? 4 - 10 cm3/s; ?T ? 100 - 200 C degrees. By applying the relevants. By applying the relevant non-dimensional parameters governing the main physical phenomena, the conclusion is that the more appropriate non-radioactive metal to simulate the atomization of U- 7 % Mo is gold

30

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)

31

Multipurpose recovery techniques of uranium-molybdenum intergrown ores at abroad  

International Nuclear Information System (INIS)

The multipurpose separation and recovery techniques of the uranium-molybdenum inter-grown ores at abroad are summarized, which include acid curing, pressure alkaline leaching as well as methods of ion exchange, extraction, active carbon absorption and precipitation at industry, and their state of research development. (authors)

32

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

33

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

34

PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY  

Energy Technology Data Exchange (ETDEWEB)

The Savannah River Site will recycle a nuclear fuel comprised of 90% uranium-10% molybdenum by weight. The process flowsheet calls for dissolution of the material in nitric acid to a uranium concentration of 15-20 g/L without the formation of precipitates. The dissolution will be followed by separation of uranium from molybdenum using solvent extraction with 7.5% tributylphosphate in n-paraffin. Testing with the fuel validated dissolution and solubility data reported in the literature. Batch distribution coefficient measurements were performed for the extraction, strip and wash stages with particular focus on the distribution of molybdenum.

Pierce, R; Ann Visser, A; James Laurinat, J

2007-10-15

35

Spectrographic analysis of uranium-molybdenum alloys; Analisis espectrografico de aleaciones uranio-molibdeno  

Energy Technology Data Exchange (ETDEWEB)

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 % MoO{sub 3}. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO{sub 3}. (Author) 5 refs.

Roca, M.

1967-07-01

36

SASSE MODELING OF A URANIUM MOLYBDENUM SEPARATION FLOWSHEET  

Energy Technology Data Exchange (ETDEWEB)

H-Canyon Engineering (HCE) is evaluating the feasibility of processing material from the Super Kukla Prompt Burst Reactor, which operated at the Nevada Test Site from 1964 to 1978. This material is comprised of 90 wt % uranium (U) (at approximately 20% 235U enrichment) alloyed with 10 wt % molybdenum (Mo). The objective is to dissolve the material in nitric acid (HNO{sub 3}) in the H-Canyon dissolvers and then to process the dissolved material through H-Canyon First and Second Cycle solvent extraction. The U product from Second Cycle will be sent to the highly enriched uranium (HEU) blend down program. In the blend down program, enriched U from the 1EU product stream will be blended with natural U at a ratio of 1 part enriched U per 3.5 parts natural U to meet a reactor fuel specification of 4.95% 235U before being shipped for use by the Tennessee Valley Authority (TVA) in its nuclear plants. The TVA specification calls for <200 mg Mo/g U (200 ppm). Since natural U has about 10 mg Mo/g U, the required purity of the 1EU product prior to blending is about 800 mg Mo/g U, allowing for uncertainties. HCE requested that the Savannah River National Laboratory (SRNL) define a flowsheet for the safe and efficient processing of the U-10Mo material. This report presents a computational model of the solvent extraction portion of the proposed flowsheet. The two main objectives of the computational model are to demonstrate that the Mo impurity requirement can be met and to show that the solvent feed rates in the proposed flowsheet, in particular to 1A and 1D Banks, are adequate to prevent refluxing of U and thereby ensure nuclear criticality safety. SASSE (Spreadsheet Algorithm for Stagewise Solvent Extraction), a Microsoft Excel spreadsheet that supports Argonne National Laboratory's proprietary AMUSE (Argonne Model for Universal Solvent Extraction) code, was selected to model the U/Mo separation flowsheet. SASSE spreadsheet models of H-Canyon First and Second Cycle solvent extraction show that a standard unirradiated fuel flowsheet is capable of separating U from Mo in dissolved solutions of a U/Mo alloy. The standard unirradiated fuel flowsheet is used, except for increases in solvent feed rates to prevent U refluxing and thereby ensure nuclear criticality safety and substitution of higher HNO{sub 3} concentrations for aluminum nitrate (Al(NO{sub 3})){sub 3} in the feed to 1A Bank. (Unlike Savanah River Site (SRS) fuels, the U/Mo material contains no aluminum (Al). As a result, higher HNO3 concentrations are required in the 1AF to provide the necessary salting.) The TVA limit for the final blended product is 200 {micro}g Mo/g U, which translates to approximately 800 mg Mo/g U for the Second Cycle product solution. SASSE calculations give a Mo impurity level of 4 {micro}g Mo/g U in the Second Cycle product solution, conservatively based on Mo organic-to-aqueous distributions measured during minibank testing for previous processing of Piqua reactor fuel. The calculated impurity level is slightly more than two orders of magnitude lower than the required level. The Piqua feed solution contained a significant concentration of Al(NO{sub 3}){sub 3}, which is not present in the feed solution for the proposed flowsheet. Measured distribution data indicate that, without Al(NO{sub 3}){sub 3} or other salting agents present, Mo extracts into the organic phase to a much lesser extent, so that the overall U/Mo separation is better and the Mo impurities in the Second Cycle product drop to negligible concentrations. The 1DF U concentration of 20 g/L specified by the proposed flowsheet requires an increased 1DX organic feed rate to satisfy H-Canyon Double Contingency Analysis (DCA) guidelines for the prevention of U refluxing. The ranges for the 1AX, 1BS, and 1DX organic flow rates in the proposed flowsheet are set so that the limiting ratios of organic/aqueous flow rates exactly meet the minimum values specified by the DCA.

Laurinat, J

2007-05-31

37

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

38

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

39

Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in ?(bcc) Uranium-Molybdenum Alloy  

Science.gov (United States)

U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

2013-02-01

40

Update on U-Mo monolithic and dispersion fuel development  

International Nuclear Information System (INIS)

The advanced fuel development effort for the Reduced Enrichment for Research and Test Reactors (RERTR) program is developing high-density fuel based on the uranium-molybdenum system. Fuel development has concentrated on U-Mo powder dispersed in a pure aluminum matrix. Recent questions about the fissile loading limit and the in-pile performance of this fuel type have led to intensified efforts to fabricate monolithic fuel and to modify the matrix of the dispersion fuel. These fuel types will be irradiated in upcoming tests. (author)

 
 
 
 
41

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

42

UPDATE ON MONOLITHIC FUEL FABRICATION METHODS  

International Nuclear Information System (INIS)

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

43

Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element  

International Nuclear Information System (INIS)

By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the ? ? ? transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the ? ? ? transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form ? at ordinary temperatures after quenching from the ? and ? regions. The ? phase is particularly unstable and changes into needles of the ? form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The ? phase obtained by quenching from the ? phase region is more stable than that obtained by quenching from the ? region. Chromium is a more effective stabiliser of the ? phase than is iron. Unfortunately it causes serious surface cracking. The ? ? ? transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct ? ? ? transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C/s. He has however observed the formation of several martensitic structures. (author)

44

An interdiffusional model for prediction of the interaction layer growth in the system uranium molybdenum/aluminum  

Science.gov (United States)

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U-Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U-Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U-Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.

Soba, A.; Denis, A.

2007-03-01

45

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel  

Science.gov (United States)

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 to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate 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 Zr 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 E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

46

Study and comparison of analytical methods for dosing molybdenum in uranium-molybdenum alloys; Etude et comparaison de methodes d'analyses du molybdene dans les alliages uranium - molybdene  

Energy Technology Data Exchange (ETDEWEB)

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) [French] On expose plusieurs methodes de dosage du molybdene dans les alliages uranium-molybdene par des techniques aussi diverses que: spectrophotometrie d'absorption moleculaire, spectrographie d'emission, fluorescence de rayons X, spectrophotometrie d'absorption atomique. Apres une comparaison portant sur des echantillons dont les teneurs en molybdene sont comprises entre 1 et 10 pour cent en poids, on conclut a l'interet de l'emploi de certaines des methodes exposees pour des analyses de serie. (auteur)

Buffereau, M.; Genty, C.; Houin, C.; Lavaud, M.; Leclainche, C.; Levrard, J.; Pichotin, B.; Robichet, J. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes Nucleaires

1968-07-01

47

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

48

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

International Nuclear Information System (INIS)

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.

49

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

50

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

51

Study of the transformation of uranium-niobium alloys with low niobium concentrations, tempered from the gamma and beta + gamma 1 regions and then annealed at different temperatures. Comparison with uranium-molybdenum alloys (1963)  

International Nuclear Information System (INIS)

The author shows that uranium-niobium alloys, like uranium-molybdenum alloys, tempered from the gamma region, give a martensitic phase with a structure deriving from that of alpha uranium by a slight contraction parallel to the axis [001], The critical cooling rate allowing the formation of this martensite is 80 deg. C/s at 750 deg. C. Retention of the beta phase of uranium-niobium alloys is particularly difficult, the critical retention rate being 700 deg. C/s at 668 deg. C for an alloy containing 2.5 at. per cent of Nb. This beta phase is completely converted to the alpha phase at room temperature in about 6 hours. The TTT curves of this beta alloy are effectively reduced to the lower branch of the lower 'C'. The beta phase conversion law is expressed as: 1-x = exp. (kt)n x being the degree of progression of the conversion, t the time, n an exponent no-varying with temperature and having approximately the value 2 for the alloy considered, k an increasing function of temperature. The activation energy of conversion is of the order of 14,600 cal/mole. Niobium is much less active than molybdenum as a stabiliser of beta uranium. (author)

52

The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing  

International Nuclear Information System (INIS)

Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

53

The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing  

Energy Technology Data Exchange (ETDEWEB)

Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

2001-10-01

54

Investigations of a reduced enrichment dispersion fuel (U-Mo alloy in aluminium matrix) for research reactor fuel pins  

Energy Technology Data Exchange (ETDEWEB)

Russia possesses considerable experience in utilisation of uranium-molybdenum alloys containing in dispersion fuel composition no more than 6 g/cm{sup 3} uranium. The feasibility of utilising the U-9 mass.% Mo alloy with reduced enrichment uranium (< 20%) in research reactor dispersion fuel pins has been analysed in the IPPE. Specimens with the 40 vol.% (U-9 mass. % Mo) + 60 vol.% Al fuel have been fabricated by hot pressing. Investigations of thermal physical properties of this fuel as well as tests for compatibility of U-Mo alloy with Al have been carried out in a wide temperature range. Corrosive tests of dispersion fuel have been realised in water. A flow chart of reproducing wastes from fuel pin production has been considered. The results of works carried out enable to hope on successful solution of the problem of utilisation high-density U-Mo fuel in research reactors. (author)

Aden, V.G. [Research and Development Institute of Power Engineering (RDIPE), Moscow (Russian Federation); Popov, V.V.; Rusanov, A.Ye.; Troyanov, V.M. [State Scientific Centre of Russian Federation, Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1999-07-01

55

Design and Testing of Prototypic Elements Containing Monolithic Fuel  

International Nuclear Information System (INIS)

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.

56

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)

57

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

58

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

59

Application of the DART Code for the Assessment of Advanced Fuel Behavior  

International Nuclear Information System (INIS)

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO2 fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)

60

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.

 
 
 
 
61

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

62

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)

63

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)

64

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

65

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

66

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

67

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

68

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

69

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.

70

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)

71

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

72

Fabrication and characterisation of uranium, molybdenum, chromium, niobium and aluminium  

International Nuclear Information System (INIS)

This paper describes fabrication of binary uranium alloys by melting and casting. The following alloys with nominal composition were obtained by melting in the vacuum furnace: uranium with niobium contents from 0.5%- 4.0% and uranium with molybdenum contents from 0.4% - 1.2%. Uranium alloys with chromium content from 0.4% - 1.2% and uranium alloy with 0.12% of aluminium were obtained by vacuum induction furnace (electric arc melting)

73

Thermal expansion and specific heat of uranium molybdenum alloy  

International Nuclear Information System (INIS)

Thermal expansion of uranium + 6 w/o molybdenum was measured by using high-temperature dilatometer in the temperature range of 400 K to 725 K. Specific heat was also measured for the same alloy in the temperature range of 373 K to 573 K by using TA Instrument DSC System (Model Thermal Analyst 2000). The corresponding polynomial expression for percentage thermal expansion in the temperature range of 400 K to 725 K is given as : (?L/L0) x 100 = 1.824- 0.013 T + 2.717 x 10-5T2 . (author)

74

Dispersed and monolithic plate type U-Mo nuclear fuels  

International Nuclear Information System (INIS)

The conversion of high flux reactors to LEU requires the development of monolithic high density fuels. The radiation performance of uranium-molybdenum (UMo) high density fuel has shown excellent behaviour of fission gas products retention and reasonable swelling. The development of monolithic UMo is usually approached from the point of view of an aluminium cladding. Interface interactions between these two materials could be an issue and fabrication procedures have to be radically changed because of the big difference in mechanical properties of core and clad. An alternative way of looking at these scenario is using zircalloy-4 (Zry-4) cladding for the U-Mo monolithic fuel. The basic idea was to study the possibility of colamination of both materials to fabricate plates. A series of preliminary studies where performed to look at the feasibility of this alternative. Calculations where done to analyze stress generation because of the different coefficient of thermal expansion of both materials. Diffusion couples where studied for evaluation of the interface growth. Zry-4 plates where colaminated for selecting ranges of colamination temperatures and deformation steps. Several miniplates where elaborated with depleted uranium and 20% enriched uranium (LEU) for irradiation purposes in the RERTR experiments in the Advanced Testing Reactor (ATR, INL). All miniplates where elaborated looking at scale production of full size plates for irradiation purposes and fuel fabrication irradiation purposes and fuel fabrication. The fabrication steps are the casting of the alloy, machining to core dimensions, cutting of frame and lids, welding (TIG) of boundaries, hot colamination, straightening of plates, abrasive finishing of plate surfaces and cutting to final dimensions. Characterization results will also be shown: metallography, ultrasonic scanning (scan-C), radiography, etc. The elaboration of U-Mo monolithic plates with Zry-4 cladding, as achieved, is a possible flexible technological alternative for fuel fabrication and conversion purposes that does not involve big changes in the usual equipment of production plants of plate fuels. (author)

75

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

International Nuclear Information System (INIS)

As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initialions 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.

76

Fabrication and characterization of atomized U-Mo powder dispersed fuel compacts for the RERTR-3 irradiation test  

International Nuclear Information System (INIS)

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 such as as-atomized U-10Mo, phase decomposed U-10Mo (alpha+gamma), homogenized U-10Mo, U-7Mo, U-6Mo, U-6.1Mo-0.9Ru, and U-6Mo-1.7Os. 25 fuel plates, referred to as nano-plates, were produced with atomized fuel compacts at ANL-W. The relationship between the volume fraction of fuel and the green density of the compacts was established. The relative density of the compacts increases with decreasing volume fraction of fuel powder. The compressibility of comminuted powder compacts was larger than that of the atomized powder compacts due to the fragmentation of comminuted particles. The green strength of comminuted powder compacts is higher than that of the atomized powder compact. This seems to have resulted from the smaller pore size and the larger contact area between the comminuted fuel powders and Al powders. It is suggested that the compacting condition adjustment be required to fabricate the atomized powder compacts having comparable green strength. (author)

77

A physical description of fission product behavior fuels for advanced power reactors  

International Nuclear Information System (INIS)

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO2 power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behaviorel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the ?-, intermediate- and ?-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified

78

A physical description of fission product behavior fuels for advanced power reactors.  

Energy Technology Data Exchange (ETDEWEB)

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

2007-10-18

79

Cross section generation and physics modeling in a feasibility 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 has been initiated at ORNL to examine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The current study is limited to steady-state, nominal operation and are focused on the determination of the fuel requirements, primarily density, that are required to maintain the performance of the reactor. Reactor physics analyses are reported for a uranium-molybdenum alloy that would be substituted for the current fuel - U3O8 mixed with aluminum. An LEU core design has been obtained and requires an increase in 235U loading of a factor of 1.9 over the current HEU fuel. These initial results indicate that the conversion from HEU to LEU results in a reduction of the thermal fluxes in the central flux trap region of approximately 9 % and in the outer beryllium reflector region of approximately 15%. Ongoing work is being performed to improve upon this initial design to further minimize the impact of conversion to LEU fuel. (authors)

80

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

 
 
 
 
81

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

82

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

83

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

84

? ????? ??? fuel cell ??? ???????? ?????  

Digital Repository Infrastructure Vision for European Research (DRIVER)

? ???????????? ???????? ??????? ?? ??????????? ?? ???????? ??? ?????????? ????????? ???? ?????????? ??????? ????? Fuel Cell, ?? ??? ????? ????????? ?????? ??? ???? ?????????????? ???? ?? ?????. ???? ???????????? ?????????? ??? ????? ?????? ?? ???????????? ??? ?????? ??????...

??????????, ???????

2009-01-01

85

Relation between Gamma Decomposition and Powder Formation of ?-U8Mo Nuclear Fuel Alloys via Hydrogen Embrittlement and Thermal Shock  

Directory of Open Access Journals (Sweden)

Full Text Available Gamma uranium-molybdenum alloys have been considered as the fuel phase in plate type fuel elements for material and test reactors (MTR, due to their acceptable performance under irradiation. Regarding their usage as a dispersion phase in aluminum matrix, it is necessary to convert the as cast structure into powder, and one of the techniques considered for this purpose is the hydration-dehydration (HDH. This paper shows that, under specific conditions of heating and cooling, ?-UMo fragmentation occurs in a non-reactive predominant mechanism, as shown by the curves of hydrogen absorption/desorption as a function of time and temperature. Our focus was on the experimental results presented by the addition of 8% weight molybdenum. Following the production by induction melting, samples of the alloys were thermally treated under a constant flow of hydrogen for temperatures varying from 500°C to 600°C and for times of 0.5 to 4 h. It was observed that, even without a massive hydration-dehydration process, the alloys fragmented under specific conditions of thermal treatment during the thermal shock phase of the experiments. Also, it was observed that there was a relation between absorption and the rate of gamma decomposition or the gamma phase stability of the alloy.

Fábio Branco Vaz de Oliveira

2014-10-01

86

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

Energy Technology Data Exchange (ETDEWEB)

Highlights: Black-Right-Pointing-Pointer GTRI program supports conversion from HEU to LEU. Black-Right-Pointing-Pointer High performance research reactors require a dense LEU fuel such as U-10Mo foils. Black-Right-Pointing-Pointer Dissolution conditions for U-10Mo foils in acidic media have been optimized. Black-Right-Pointing-Pointer Solvent-extraction processing can be used to recover U lost in fuel fabrication. Black-Right-Pointing-Pointer Flowsheets were developed using Argonne-design contactors but other contactors can be used as well. - Abstract: 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(NO{sub 3}){sub 3}. HNO{sub 3} and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr{sub 2} intermetallic, explosive complex, while meeting the requirements needed for further processing.

Youker, Amanda J., E-mail: youker@anl.gov [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2012-08-15

87

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

88

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

Energy Technology Data Exchange (ETDEWEB)

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 UCl{sub 3}. 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 K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.

Van Kleeck, M. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Willit, J.; Williamson, M.A. [Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Fentiman, A.W. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)

2013-07-01

89

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

90

Characterization of poly- and single-crystal uranium-molybdenum alloy thin films  

Digital Repository Infrastructure Vision for European Research (DRIVER)

Poly- and single-crystal thin films of U-Mo alloys have been grown both on glass and sapphire substrates by UHV magnetron sputtering. X-ray and Electron Backscatter Diffraction data indicate that for single-crystal U1-xMox alloys, the pure cubic uranium gamma-phase exists for x > 0.22 (10 wt.% Mo). Below 10 wt.% Mo concentration, the resulting thin film alloys exhibited a mixed alpha-gamma uranium phase composition.

Adamska, A. M.; Springell, R.; Scott, T. B.

2013-01-01

91

Study of the ductile-brittle transition in uranium-molybdenum alloys  

International Nuclear Information System (INIS)

The toughness and mechanical properties in tension and compression for binary U-Mo (Mo=6, 8, 10, 12%) and ternary U-Mo-Zr (U-Mo 10 - Zr 1) and U-Mo-Ti (U-Mo 8 - Ti 1) alloys are determined between -196 and +1400C. Empirical relations of Hahn, Hoagland and Rosenfield allow to correlate characteristics of these alloys

92

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)

93

Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content  

International Nuclear Information System (INIS)

We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the ?-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The ? grain is fine, the ?-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the ?-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the ?-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors)

94

Extraction separation of thorium from uranium, molybdenum and other elements in carbonate solutions  

International Nuclear Information System (INIS)

Extraction of thorium carbonate complexes from various solvents by primary, secondary and tertiary alkylamine solutions and by salts of quaternary ammonium bases has been studied. The most complete extraction is provided by primary alkyl amines. At pH=8 the (C12H25NH3)4 [Th(CO3)4(H2O)sub(x)] associate is extracted. Results on extraction separation of thorium from uranium (6), zirconium, molybdenum, cobalt and magnesium in carbonate solutions by primary alkyl amine solutions in kerosene are reported

95

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

Energy Technology Data Exchange (ETDEWEB)

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.

Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

2010-02-01

96

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly in which a plurality of UO2 fuel rods and a plurality of MOX fuel rods are arranged in parallel with each other and each of the fuel rods is supported by upper and lower tie plates, a predetermined number of MOX fuel rods are secured to the upper and the lower tie plates to provide fastened fuel rods. Further, the MOX fuel rods and the UO2 fuel rods except for the fastened fuel rods are supported movably to the upper and the lower tie plates to provide standard fuel rods. With such a constitution, elongation of the MOX fuel rods as the standard fuel rods is not restricted. Accordingly, distortion of the fuel rod caused by the difference of elongation of the MOX fuel rod and the UO2 fuel rod can be certainly prevented, to maintain fuel rod integrity. (T.M.)

97

Fuel assembly  

International Nuclear Information System (INIS)

A fuel assembly for a BWR type reactor comprises a plurality of fuel rods and at least one water rod arranged in an n x n matrix. n is 9 or greater, for example, 9. 74 fuel rods are disposed except for the space for the water rods of the 9 x 9 matrix. Among them, 60 fuel rods are MOX fuel rods filled with uranium-plutonium mixed oxide fuels, and the residual 14 fuel rods are uranium-gadolinium fuel rods filled with uranium oxide fuels to which gadolinium is added as a burnable poison. The ratio of the fission plutonium contained in the MOX fuel rods based on the total fuel rods in the fuel assembly is determined to 3.4% by weight or less. The MOX fuel rods comprise two kinds of fuel rods having different plutonium enrichment degrees. (I.N.)

98

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To ensure the fuel assembly integrity without imparing the reliability of uranium-plutonium fuel rods even if the effective utilization of plutonium is intended. Constitution: In a fuel assembly comprising a plurality of bundled uranium-plutonium fuel rods (MOX fuel rods), the pressurizing degree for helium gases sealed in MOX fuel rods is made greater than that for helium gases sealed in uranium fuel rods. Accordingly, if the Pu enrichment degree is increased in the MOX fuel rods, the temperature upon burning the fuel pellets in the MOX fuel rods is substantially the same as the temperature upon burning of fuel pellets in the uranium fuel rods. As a result, integrity of the MOX fuel assembly can be ensured. (Takahashi, M.)

99

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly having MOX fuel rods charged with MOX fuel pellets and UO2 fuel rods charged with ordinary UO2 pellets, a fuel can having crystal bar zirconium of excellent PCI resistant property lined on the inner surface is used for the UO2 fuel rod, and a fuel can having economical sponge zirconium of excellent mass productivity lined on the inner surface is used for the MOX fuel rod. Accordingly, a fuel assembly excellent in the performance and also from an economical point of view can be attained. (T.M.)

100

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

 
 
 
 
101

Fuel assembly  

International Nuclear Information System (INIS)

The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ? 173m, Y ? - 9.7X + 292, Y ? - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

102

Fuel assembly  

Energy Technology Data Exchange (ETDEWEB)

The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y {>=} 173m, Y {<=} - 9.7X + 292, Y {<=} - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

1998-02-20

103

Fuel assembly  

International Nuclear Information System (INIS)

In a BWR type fuel assembly, a volume ratio of a plenum portion relative to a fuel portion is selected within a range from 0.11 to 0.15. Alternatively, a length ratio of the plenum portion relative to the fuel portion is selected within a range from 0.12 to 0.16. In the thus constituted fuel assembly, since a greater plenum portion is ensured for a MOX fuel rod than for an uranium fuel rod, inner pressure of the MOX fuel rod is maintained substantially identical with that of the uranium fuel rod even if greater amount of FP gases are released from the MOX fuel than the uranium fuel. Accordingly, risk of causing excess stresses to cladding tubes which would lead to fuel rupture can be remarkable reduced. (T.M.)

104

Fuel assembly  

International Nuclear Information System (INIS)

The present invention concerns a fuel assembly in a light water cooled reactor, and it provides a fuel assembly suitable to plutonium containing fuels. That is, local power increase is avoided without increasing the ratio of natural uranium in an MOX fuel pellet. Specifically, the outer diameter of the MOX fuel pellet is decreased than that of a usual uranium fuel pellet. This can decrease the volume of the MOS fuel pellet, thereby enabling to avoid power increase of the MOX fuel pellet. It is because the power of the fuel pellet is in proportion to the volume and the power density of the fuel pellet. As a result, local power increase can be avoided to ensure thermal margin without increasing the ratio of natural uranium in the MOX fuel. (I.S.)

105

Fuel and nuclear fuel cycle  

International Nuclear Information System (INIS)

The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

106

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.)

107

Fuel assembly and fuel rod  

International Nuclear Information System (INIS)

Fuel assemblies comprising only UO2 fuel rods can be assembled and tested by direct contact with the UO2 fuel rods and the fuel assembly. However, since fuel assemblies containing MOX fuel rods have high irradiation dose caused by ? decay of Pu-241 having short half-life, operator's access is impossible, and the fuel assemblies are assembled automatically or by remote operation. Inspections after assembling the fuel assemblies are also performed automatically or by remote handling. Then, the shape of the upper end plug of the UO2 fuel rods and that of MOX fuel rods are made optically different such that the UO2 fuel rods and the MOX fuel rods can be distinguished by measuring or analyzing reflected light reflected from the upper end plug of the fuel rod when light is irradiated from a position above the fuel assembly. Inspection for the fuel assembly comprising mixed UO2 MOX fuel rods which has to be done automatically or by remote operation after assembling the fuel assemblies can be facilitated. (N.H.)

108

Fuel assembly  

International Nuclear Information System (INIS)

It is considered that pellet-cladding tube mechanical interaction (PCI) characteristics become severe in an UO2 fuel rod compared with a MOX fuel rod since creep of a pellet is small and there is a possibility of undergoing a large power fluctuation range. Then, in a fuel assembly having UO2 fuel rods and MOX fuel rods arranged in a lattice-like configuration, zirconium of excellent PCI characteristics is lined on the inner side of the cladding tube only for the UO2 fuel rod in the present invention. Then, a cladding tube made of zirconium alloy excellent in massproduction is used for the MOX fuel rod. (T.M.)

109

Fuel element  

International Nuclear Information System (INIS)

The present invention concerns a fuel element of a reactor. As the fuel element, a fuel can in which a pure zirconium metal layer is metallurgically bonded at the inner surface of a fuel can made of a zirconium alloy is used. A plurality of fuel pellets closely coated at each of their surfaces with a pure thin zirconium layer are charged in the fuel can. A plenum for storing gases is formed in the upper portion of the fuel pellets. A spring is disposed in the plenum for preventing vertical movement of the fuel pellets and the upper and lower ends of the fuel can are sealed with plugs. With such a constitution, transfer of oxygen from the fuel pellets to the pure zirconium metal layer at the inner surface of fuel cans can be completely prevented. Accordingly, the pure zirconium metal layer at the inner surface of the fuel can is prevented from embrittlement and, even if mechanical interactions should be caused between the fuel pellets and the fuel can, stress corrosion cracks of the fuel can are not caused. (I.N.)

110

Fuel assembly  

International Nuclear Information System (INIS)

A fuel assembly comprises a plurality of fuel rods arranged in a lattice. The plurality of fuel rods comprise longer fuel rods and shorter fuel rods having a shorter effective portion than that of the longer fuel rod. The effective portion of the shorter fuel rod is divided into upper and lower two regions. Assuming the concentration of fission material of fuel pellets filled in the lower region thereof as A, and the concentration of the fissile material of the fuel pellets filled in the upper region as B, the relation B ? A is defined. In this case, A is smaller than the average concentration of the fissile materials in a cross section of the assembly. The concentration of the fissile material of the fuel pellets filled in the effective portion of the longer fuel rod is uniform in a vertical direction. With such a constitution, the plenum of the shorter fuel rod is made shorter than the plenum of the longer fuel rod. Accordingly, improvement of fuel economy and reactor shut down margin due to increase of fuel inventories can be attained. (I.N.)

111

Fuel assembly  

International Nuclear Information System (INIS)

This invention concerns a fuel assembly improved with neutron utilization factor and provided with sufficient reactor shutdown margin. Upon constituting a fuel assembly by disposing short-sized fuel rods and usual fuel rods mixed together, a portion of the gas plenum region for a short-sized fuel rod is made finer than the fuel rod diameter in a fuel pellet region. In the fuel assembly comprising such short-sized fuel rods having the gas plenum portion and the long-sized fuel rods, light water can be present also in the space which has been the gas plenum portion in the conventional short-sized fuel rod. Accordingly, since the water/fuel volume ratio in the upper portion of the fuel assembly is increased, the reactor shutdown margin is increased. Thus, the pellet region in the short-sized fuel rod can be made more longer than usual while maintaining a required reactor shutdown margin. As a result, since the uranium charging amount can be increased for the entire assembly, the fuel economy can be improved. (K.M.)

112

Fuel assembly  

International Nuclear Information System (INIS)

The ratio t/D between the wall thickness t and the outer diameter D of a fuel rod cladding tube in a fuel assembly is increased in an MOX fuel rod compared with an uranium fuel rod, to constitute a fuel assembly as a structure which is strong also mechanically. With such a constitution, even if gaseous fission products are released from the MOX fuel pellet due to high burnup degree and the inner pressure of the MOX fuel rod is increased, integrity of the cladding tube can be kept. Accordingly, since the possibility of losing integrity of the MOX fuel rod can be reduced even if high burnup degree is attained, the reactor can be operated safely. Further, since a fuel assembly capable of attaining high burnup degree can be provided, economical property can be improved. (T.M.)

113

Fuel assembly  

International Nuclear Information System (INIS)

An MOX fuel pellet used for an island type MOX fuel has lower heat conductivity than a UO2 pellet of an UO2 fuel rod. Accordingly, if fuel rods having the same linear power density are compared, the highest pellet temperature of the MOX fuel rod is higher than that of a UO2 fuel rod. As the pellet temperature is elevated, the amount of gaseous fission products (FP gas) released from the pellet to the inner side of the fuel rod is increased, to increase the pressure in the fuel rod. Then, according to the present invention, in the MOX fuel assembly having the MOX fuel rods, the gap between the pellet and the cladding tube of the MOX fuel rod is made smaller than that between the pellet and the cladding tube of the UO2 fuel rod. This can improve the heat transfer coefficient of the gap between the pellet and the cladding tube, thereby enabling to lower the highest temperature of the pellet. Accordingly, the fuel rod integrity in the MOX fuel assembly can be improved. (T.M.)

114

Fuel rod  

International Nuclear Information System (INIS)

In the fuel rod with UC2-PuC2 fuel with 93% U235 enrichment and 15% Pu as well as a plenum, an auxiliary gas or gas mixture is filled into the plenum. The auxiliary gas (in addition to the fission gases) of He, Ne, Ar, N has a pressure between 1 and 12 atmospheres and a higher thermal conductivity than the fission gases. This way, heat transfer between fuel und fuel can is improved and surface deformations of the fuel can (e.g. due to swelling of the fuel) are avaided. (DG)

115

Fuel assembly  

International Nuclear Information System (INIS)

Five p-fuel rods are disposed between two large-diameter moderator rods. Ordinary fuel rods are disposed in a 9 x 9 matrix in regions other than the above-mentioned region. The ordinary fuel rod is formed by sealing fuel pellets in a fuel cladding tube, and has a gas plenum in the upper portion. The p-fuel rods are short fuel rods, and the entire length corresponds approximately to 3/4H of the entire fuel effective length H of the ordinary fuel rod. A gas plenum is formed on the lower end and an auxiliary gas plenum is formed on the upper end by way of a power spike inhibitor, and a non-fuel portion is situated thereabove. The p-fuel rods are not filled with fission nuclides at positions including portions of from 2/3H to 5/6H where reactor shut down margin is severe. With such a constitution, there can be obtained fuel assemblies constituting the reactor core of a BWR type reactor in which the reactor shut down is enabled even if the enrichment degree of fuels is increased and axial power distribution is improved. (I.N.)

116

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

117

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

118

Fuel rod  

International Nuclear Information System (INIS)

Purpose: To prevent fuel element failures by moderating the fuel-cladding interactions in fuel rods for BWR type reactors. Constitution: Fuel pellets of uranium dioxide UO sub(2 + x) for 0 < x <= 0.25 are charged in lamination within a cladding tube. Further, carbon-dioxide-absorbing substances such as magnesium oxide or calcium oxide are charged in the plenum within the cladding tube. (J.P.N.)

119

Fuel element  

International Nuclear Information System (INIS)

Purpose: To eliminate occurrence of stress corrosion crack at a fuel can by restricting the discharge amount of a corrosive fission product stored in a fuel pellet. Constitution: Small amounts of metal oxides such as calcium oxide, cobalt oxide, copper oxide, nickel oxide, chromium oxide and iron oxide having two or three valence are added to a fuel element of uranium dioxide sealed in a zirconium alloy fuel can. (Aizawa, K.)

120

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To effectively utilize gadolinia reactivity without deteriorating the thermodynamic properties of fuel in a BWR type reactor fuel element containing gadolinia as burnable poison. Constitution: One or both of isotopes Gd-155 and Gd-157 having large reactivity in gadolinia is concentrated, and a fuel element is fabricated with the concentrated isotope. (Kamimura, M.)

 
 
 
 
121

Fossil Fuels.  

Science.gov (United States)

This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

Crank, Ron

122

Nuclear fuels  

International Nuclear Information System (INIS)

Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO2gs); In-reactor behavior of UO2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO2 and MOX ceramics (Chromium oxide-doped UO2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evoluti

123

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

124

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

125

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)

126

Fuel assembly  

International Nuclear Information System (INIS)

Although shorter fuel rods have been used for increasing the burnup degree of fuels, this brings about a problem that the burnup degree of fuel rods around a shorter fuel rod is reduced. In view of the above, in the present invention, the height of a gas plenum situated in the upper end of a second fuel rod having a shorter axial length than that of the first fuel rod is made lwoer than the height of the gas plenum situated in the upper poriton of the first fuel rod. Then, the volume of a spring disposed in the gas plenum of the second fuel rod is made smaller than that of the spring disposed in the gas plenum of the first fuel rod. Further, the spring in the second gas plenum is made of a Zr-Nb alloy with less thermal neutorn cross section. Thus, it is possible to reduce wasteful absorption of neutrons in the upper end portion of the shorter fuel rod thereby improving the neutron economy. (T.K.)

127

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

Energy Technology Data Exchange (ETDEWEB)

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.

C.R. Clark; B.R. Muntifering; J.F. Jue

2007-09-01

128

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To increase the burnup degree of fuel and to increase the surplus reactivity in the latter stage of burnup. Constitution: One or more control pins having a cladding tube filled with a burnable poison (such as gadolinium 155, 157 and boron 10), thorium, and the like are incorporated into a channel in place of a part of fuel pins in the fuel assembly. The thus formed fuel assembly is restrained in reaction by the strong neutron absorbing action of the burnable poison in the early stage of burnup, as a consequence of which in the early stage of burnup, the surplus reactivity is low and extra fuel may be loaded by that portion to thereby increase the burnup degree of fuel. (Furukawa, Y.)

129

Fuel assembly  

International Nuclear Information System (INIS)

A plurality of fuel pellets are filled in a cladding tube of a fuel rod. A plenum portion is disposed above the fuel pellets. A getter tube and a spring are inserted to the plenum portion. In addition, a neutron control portion is disposed below the getter tube in the plenum portion. The neutron controlling portion is composed of a high energy neutron moderator or a neutron absorber, for example, ZrH2. Neutrons generated by nuclear fission in the fuel pellets are moderated at first in the fuel pellets. Neutrons directing in the axial direction of the fuel rod are moderated by the neutron controlling portion, further moderated by the cladding tube or an upper end plug, and then discharged into coolants. This can reduce dose equivalent rate in turbine buildings and at the boundary of sites. (I.N.)

130

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

131

Fuel cycle  

International Nuclear Information System (INIS)

The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.)

132

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

133

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.)

134

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly of a BWR type reactor, lattice space is formed to lattice frame constituting fuel spacer to which fuel rods are to be inserted. Protrusions are formed so as to surround the periphery of the fuel rod. The protrusions are turned up by forming cutting to the lattice frame and bent in the direction opposite to each other toward fuel rods. The fuel rods are disposed to the lattice frame and elastically supported by the support comprising a stopper and a spring. The protrusions disposed to the lattice frame are present in coolant flow channel so as to surround the periphery of the surface of the fuel rods. This promotes the flow of steams directing to the surface of the fuel rods, the coolants flowing in the coolant flow channel, especially liquid droplets flowing along the flow of the steam are deposited to liquid membranes which rise along the furl rods efficiently. Accordingly, the thickness of the liquid membrane is increased thereby improving the limit power of the fuel assembly. (I.N.)

135

Fuel assembly  

Energy Technology Data Exchange (ETDEWEB)

In a fuel assembly of a BWR type reactor, lattice space is formed to lattice frame constituting fuel spacer to which fuel rods are to be inserted. Protrusions are formed so as to surround the periphery of the fuel rod. The protrusions are turned up by forming cutting to the lattice frame and bent in the direction opposite to each other toward fuel rods. The fuel rods are disposed to the lattice frame and elastically supported by the support comprising a stopper and a spring. The protrusions disposed to the lattice frame are present in coolant flow channel so as to surround the periphery of the surface of the fuel rods. This promotes the flow of steams directing to the surface of the fuel rods, the coolants flowing in the coolant flow channel, especially liquid droplets flowing along the flow of the steam are deposited to liquid membranes which rise along the furl rods efficiently. Accordingly, the thickness of the liquid membrane is increased thereby improving the limit power of the fuel assembly. (I.N.)

Orii, Akihito; Kawasaki, Terufumi; Nishida, Koji; Kanazawa, Toru; Kashiwai, Shin-ichi; Nagayoshi, Takuji; Chaki, Masao

1996-06-21

136

Fuel assembly  

International Nuclear Information System (INIS)

A boiling water reactor fuel assembly is provided with an elongated, vertical stiffening device having four stiffening wings which are each attached to a wall of the fuel box. Each stiffening wing has at least one vertically directed passageway for water

137

Fuel elements  

International Nuclear Information System (INIS)

Purpose: To suppress the FP amount diffusion to the primary coolant circuit to a lower value a higher working temperature, by moderating the restriction for the highest working temperature of coated fuel particles in high temperature gas reactor fuels. Constitution: FP diffusion protection walls made of silicon carbide are disposed to the inner surface, outer surface or intermediate portion of fuel compact-containing graphite sleeves to form layers surrounding the fuel compacts. Zirconium carbide may be used in place of silicon carbide. Since the temperature at the inner and the outer surfaces of the graphite sleeves containing the fuel compacts is lower by about 200 - 2500C as compared with the temperature of the coated fuel particles as center of the fuel elements in this structure, if the fuel particles are used at a temperature of 16000C or slightly higher, the temperature at the protection walls formed to the graphite sleeve is lower than 16000C and effective FP diffusion protective effect can be maintained. (Moriyama, K.)

138

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

139

Fuel reprocessing  

International Nuclear Information System (INIS)

The reprocessing of UO2 fuels from water cooled reactors is described. The process used (PUREX process) consists to dissolve the fuel in nitric acid, to separate the nitrates by solvent extraction with TBP diluted in dodecane and to purify U and Pu. The operation scheme is given. The processing of the wastes produced (fission products, gaseous, liquid and solid wastes) is presented. The cost of the reprocessing is analysed

140

Fuel assembly  

International Nuclear Information System (INIS)

Purpose: To attain efficient fuel burning by providing a sufficient controlling performance for the reactivity at the initial stage in a BWR type reactor and prevent wasteful consumption of neutrons by gadolinium at the final stage of burning. Constitution: A fuel assembly of nuclear fuel rods for use in a BWR type reactor, for example, in a 8 column x 8 row arrangement, comprises 1 - 2 water rods through which water flows and about 8 fuel rods in which a gadolinium isotope composition containing 50 - 90 % of gadolinium 157 is contained as burnable poison (neutron absorbing burnable substance) in uranium dioxide. In such a structure, gadolinium 157 with a high neutron absorption cross section, can effectively suppress the excess reactivity of uranium dioxide fuels at the initial stage of the reactor burning, and gadolinium 157 is converted into gadolinium 158, with lower absorption cross section, does not wastefully consume neutrons, and thus the consumption of uranium dioxide can be economized. In addition, the amount of gadolinium to be incorporated can be decreased than usual, and so the worsening can be moderated of the heat conductivity of the fuel rod. (Seki, T.)

 
 
 
 
141

Fuel assembly  

International Nuclear Information System (INIS)

In a fuel assembly for use in BWR type reactors, bending of a channel box at the circumference of the reactor core is convexed toward the center of the reactor core. The channel box for the fuel assembly at the outermost circumference bends toward the control rod, while the channel box for the fuel assembly in the second row bends in the direction aparting from the control rod. In the conventional fuel assembly, cooling apertures disposed on the side of the lower tie plate for cooling incore detectors are present only on the side not facing the control rod and, thus, lack in the symmetry. Therefore, when a fuel assembly is displaced in parallel to the adjacent portion, the incore detector can not be cooled to possibly result in worsending of the insertion property of the control rod by the interference due to bending. In view of the above in the present invention, cooling apertures are formed to all four sides of the lower tie plate. Thus, the fuel assembly can be displaced in parallel to the adjacent row, thereby enabling to prevent interference due to the bending of the channel box and to use a channel box restricted by the deformation for a longer period of time. (N.H.)

142

Fuel channel for fuel assemblies  

International Nuclear Information System (INIS)

Purpose: To completely prevent a channel spacer from rotational displacement by securing the channel spacer to a fuel channel body and connecting them to each other by rotation-preventive pin. Constitution: A channel spacer is fastened at the outer side near the top end of a fuel channel body with a rivet or bolt of which the opening end is welded. A rotation-preventive pin is passed through them at any desired position excluding dimple of the fuel channel body, vicinity of rivet and the outer circumferential tapered portion on the channel spacer. The pin is welded at one end to the channel spacer. (J.P.N.)

143

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.)

144

Nuclear fuel storage arrangement  

International Nuclear Information System (INIS)

An arrangement is disclosed for the storage of nuclear reactor fuel assemblies having a section wherein fuel is present and a section wherein fuel is not present. The fuel assemblies are placed in a plurality of elongated cells which are joined together to form a cellular structure. The fuel assemblies are placed within the cells at different elevations so that the fuel-containing section of one fuel assembly is next to the non-fuel-containing sections of each fuel assembly surrounding the first fuel assembly. The vertical staggering of the fuel-containing sections achieves space reductions while maintaining the stored fuel in a subcritical assemblage. 12 claims, 10 figures

145

fuel rod  

International Nuclear Information System (INIS)

Purpose: To improve the fuel rod soundness by providing radial apertures in the bottom of an upper end plug in contact with a plenum spring to thereby prevent fusion between the plenum spring and the upper end plug upon welding. Constitution: An upper end plug in contact with a plenum spring composed of a loop spring and a coil spring for urging pellets is provided at the bottom thereof with a radial aperture in one direction or two radial apertures crossing to each other. This reduces the radial cross sectional area of the upper end plug to decrease the heat transfer from the upper end plug to the plenum spring upon welding the upper end plug to a fuel can. Thus, the temperature of the portion of the end plug in contact with the plenum spring is lowered to prevent the fusion between the upper end plug and the plenum spring, whereby the soundness of the fuel rod can be improved. (Seki, T.)

146

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

147

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

148

Fuels characterization studies. [jet fuels  

Science.gov (United States)

Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

Seng, G. T.; Antoine, A. C.; Flores, F. J.

1980-01-01

149

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.

150

Fuel Removal-Weddell  

Science.gov (United States)

... of unusable fuel: ? Six hundred fifty (650) drums of JP-4 (a kerosene-based fuel), contaminated with ... The remainder of the fuel was provided by the Russians. All of the fuel was procured in Uruguay from ...

151

Fuel oil poisoning  

Science.gov (United States)

Fuel oil poisoning occurs when someone swallows, breathes in (inhales), or touches fuel oil. This is for information only and not ... Fuel oil Kerosene Note: This list may not include all sources of fuel oil.

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.

Development, Us D.

153

Fuel element  

International Nuclear Information System (INIS)

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

154

Solid TRU fuels and fuel cycle technology  

International Nuclear Information System (INIS)

Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

155

Fuel elements in the open fuel cycle  

International Nuclear Information System (INIS)

Current needs for nuclear fuel as well as problems related to safe storage of spent fuel and/or reprocessing initiated improved technology development for fabrication of MOX fuel. From both economic and technology reasons this type of fuel seems to be attractive for revival of related research and development programs in our country

156

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

157

Fuel rod for nuclear fuel assembly  

International Nuclear Information System (INIS)

A cylindrical member made of a metal is disposed between fuel pellets and a plenum spring. The cylindrical metal member is made of the same material as a fuel cladding tube or a material having a smaller linear expansion coefficient than that of the cladding tube. If high pressure is exerted on a fuel rod, the fuel cladding tube is deformed at a plenum portion in which the plenum spring is disposed. The deformation is caused at the corner of the cylindrical metal member. Since the cylindrical metal member is softer than fuel pellets made of ceramics, there is no worry of breaking the fuel cladding tube at the corner. In addition, the corner of the cylindrical metal member is chamfered. Even if high pressure is exerted on the fuel rod, the fuel pellets can be sealed in the inside of the fuel cladding tube. (I.N.)

158

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.

2012-08-30

159

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.)

160

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

 
 
 
 
161

Fuel pellet loading apparatus  

International Nuclear Information System (INIS)

Apparatus is described for loading a predetermined amount of nuclear fuel pellets into nuclear fuel elements and particularly for the automatic loading of fuel pellets from within a sealed compartment. (author)

162

Nuclear reactor fuel elements  

International Nuclear Information System (INIS)

An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

163

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.

164

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

165

RERTR-13 Irradiation Summary Report  

International Nuclear Information System (INIS)

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. This 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.

166

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)  

International Nuclear Information System (INIS)

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable ? (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

167

HTGR reactor fuel  

International Nuclear Information System (INIS)

A critical review is given of data on the physical, chemical end other relevant properties of fuel materials and fuel elements for HTGR reactors. Available technologies for coated particles and fuel elements fabrication and appropriate quality control are described. In-core fuel behavior and spent fuel reprocessing are briefly discussed. Basic information about uranium and thorium fuel cycles is included. (author). 17 figs., 21 tabs., 87 refs

168

Dispersion fuel element fabrication  

International Nuclear Information System (INIS)

Peculiarities of dispersion fuel element fabrication are discussed. Methods of fuel rod fabrication, cladding and fuel element sealing are considered. Technology of fuel element fabrication by the methods of melting and casting, powder metallurgy is described in detail along with the methods of metallic, ceramic, and graphite coating on fuel microparticles. It is noted that in the process of fuel element fabrication emphasis is placed upon technological control of their quality

169

CANDU fuel performance  

International Nuclear Information System (INIS)

CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

170

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

171

Effect of hydrocarbon fuel type on fuel  

Science.gov (United States)

A modified jet fuel thermal oxidation tester (JFTOT) procedure was used to evaluate deposit and sediment formation for four pure hydrocarbon fuels over the temperature range 150 to 450 C in 316-stainless-steel heater tubes. Fuel types were a normal alkane, an alkene, a naphthene, and an aromatic. Each fuel exhibited certain distinctive deposit and sediment formation characteristics. The effect of aluminum and 316-stainless-steel heater tube surfaces on deposit formation for the fuel n-decane over the same temperature range was investigated. Results showed that an aluminum surface had lower deposit formation rates at all temperatures investigated. By using a modified JFTOT procedure the thermal stability of four pure hydrocarbon fuels and two practical fuels (Jet A and home heating oil no. 2) was rated on the basis of their breakpoint temperatures. Results indicate that this method could be used to rate thermal stability for a series of fuels.

Wong, E. L.; Bittker, D. A.

1982-01-01

172

Fuel pellet loading apparatus  

International Nuclear Information System (INIS)

Automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element. (auth)

173

Nuclear fuel element  

International Nuclear Information System (INIS)

Purpose: To reduce the probability of stress corrosion cracks in a zirconium alloy fuel can even when tensile stresses are resulted to the fuel can. Constitution: Sintered nuclear fuel pellets composed of uranium dioxide or a solid solution of gadolinium as a burnable poison in uranium dioxide are charged in a tightly sealed zirconium alloy fuel can. The nuclear fuel pellets for the nuclear fuel element are heat-treated in a gas mixture of carbon dioxide and carbon monoxide. Further, a charging gas containing a mixture of carbon dioxide and carbon monoxide is charged within a zirconium alloy fuel can packed with the nuclear fuel pellets and tightly sealed. (Aizawa, K.)

174

Fuel cell  

International Nuclear Information System (INIS)

The present invention concerns a fuel cell using depleted uranium. A hydrogen occluding alloy comprising a depleted uranium is utilized. The hydrogen occluding alloy is contaminated by radioactive materials. A great amount of hydrogen is stored in the hydrogen occluding alloy by ? rays from the radioactive materials. On the other hand, hydrogen is easily released from the hydrogen occluding alloy at a relatively low heating temperature by ? rays. If water is electrolyzed by using a surplus electric power or the like, a great amount of hydrogen is generated and the hydrogen can be easily stored, and oxygen generated simultaneously can also be stored. Hydrogen and oxygen are reacted in a reactor, and electrons are taken out from the electrode to generate AC power by way of a DC/AC convertor. (N.H.)

175

Fuel storage  

International Nuclear Information System (INIS)

ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well knownjor challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

176

Nuclear fuel cycle  

International Nuclear Information System (INIS)

A sphere-pack type fuel pin comprising nuclear fuels comprising a uranium/plutonium mixed nitride fuel and a minor actinoid nitride fuel, and a sodium thermal bond material is prepared. The minor actinoid element includes, for example, neptunium 237, americium 243 and curium 247. The fuel pins are loaded to a reactor core, and the fuels are burnt. Spent fuels are put to molten salt electrolysis. Uranium, plutonium and minor actinoids deposited on a cathode are converted to higher order nitrides, and they are recovered. The recovered higher order nitrides are converted to mononitrides. Then, a nuclear fuel comprising a uranium/plutonium mixed nitride fuel and a minor actinoid nitride fuel are manufactured from the mononitride. This can ensure inherent reactor safety. In addition, short fuel multiplication time and high uranium resource utilization ratio can be attained. (I.N.)

177

Nuclear fuel activities in Canada  

International Nuclear Information System (INIS)

Nuclear fuel activities in Canada are considered in the presentation on the following directions: Canadian utility fuel performance; CANDU owner's group fuel programs; AECL advanced fuel program (high burnup fuel behaviour and development); Pu dispositioning (MOX) activities. 1 tab

178

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

179

Spent fuel management strategies  

International Nuclear Information System (INIS)

Nuclear fuel cycle is divided into two sections; front end and back end of the fuel cycle. Front end of the fuel cycle, which covers all the activities of the fuel cycle before the fuel goes into the reactor has better developed and well-defined technologies. For storage of the spent fuel which are subjects of the back end of the fuel cycle, the waste management policies are not so well defined. There are three approaches that exist today for management of spent fuel. 1. For once through or open fuel cycles direct disposal of spent fuel in a deep geological repository, 2. For closed fuel cycles reprocessing of spent fuel and recycling of the recovered plutonium and uranium in new mixed oxide (MOX) fuels, 3. The spent fuel is placed in long term interim storage pending a decision as to its ultimate reprocessing or disposal. There are so large scale geological repositories for the final disposal of spent fuel in operation. Studies on suitable site selection, design, construction and licensing take about 30-40 years. Reprocessing, on the other hand, produces plutonium and is therefore under close inspection because of the Non Proliferation Treaty. Today more countries are delaying their final decision about the spent fuel management approach and using the long term interim storage approach

180

Integrated fuel processor development  

International Nuclear Information System (INIS)

The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-boarome in order to develop practical, on-board devices are discussed

 
 
 
 
181

Nuclear reactor fuel element  

International Nuclear Information System (INIS)

In order to ensure that the fuel rods of a fuel element bundle are protected against vibration and deformation, spacers in the form of at least three wires are provided, which are arranged spirally around a fuel rod. The fuel elements for fast nuclear reactors are envisaged here. (TK)

182

Engine fuel control system  

Energy Technology Data Exchange (ETDEWEB)

This patent describes a fuel control system for an internal combustion engine, in which the idle fuel quantity is adjusted to maintain a predetermined desired idle speed, and in which the transition between the idle fuel quantitiy and a scheduled off-idle fuel quantity is blended to provide for smooth engine operation.

Ament, F.; Peden, R.A.; Ausen, J.E.

1988-01-05

183

Nuclear fuel rod  

International Nuclear Information System (INIS)

An improvement for the construction of a fuel rod for water cooled nuclear reactors is proposed, which should prevent the fuel cans made of zirconium or niobium alloys from reacting even if water enters (due to a fuel can fracture). It is proposed that the internal surface of the fuel can should be covered with an oxide layer at least 2 ?m thick. (UWI)

184

Bulk Fuel Man.  

Science.gov (United States)

This student guide, one of a series of correspondence training courses designed to improve the job performance of members of the Marine Corps, deals with the skills needed by bulk fuel workers. Addressed in the four individual units of the course are the following topics: bulk fuel equipment, bulk fuel systems, procedures for handling fuels, and…

Marine Corps Inst., Washington, DC.

185

Fuel cell  

Energy Technology Data Exchange (ETDEWEB)

Since a temperature of a fuel cell rises due to its reaction heat during its operation, it is cooled in order to maintain the specific actuation temperature. For this purpose, a straightening vane is provided in an outlet manifold so that the cooling gas may be distributed uniformly above and below the cell stack. Heretofore, the above straightening vane has been fixed and done no change of the flow distribution during the time of temperature rising and the time of operation, consequently the cell temperature has not been uniform vertically. In this invention, the above straightening vane in the manifold is composed of a shape memory alloy which has its transformation temperatures between the inlet temperature of the cooling gas during the time of operation and the inlet temperature of the heating gas during the time of temperature rising so that the gas distribution to the upper and lower terminals of the cell may become less during the time of operation and more during the time of temperature rising, resulting in the uniform temperature throughout above and below the cell stack in both stages of operation and temperature rising. (2 figs)

Saito, Rokuya; Tajima, Osamu; Shindo, Koji; Kumiya, Masaichi

1987-10-17

186

MOX fuel assembly design  

International Nuclear Information System (INIS)

This patent describes improvement in a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; the improvement comprises the fissile material including a mixture of uranium and recovered plutonium in rods of the fuel bundle at locations other than the corners of the fuel bundle; and, neutron absorbing material being located in rods of the fuel bundle at rod locations adjacent the corners of the fuel bundles whereby the neutron absorbing material has decreased shielding from the plutonium and maximum exposure to thermal neutrons for shaping the cold reactivity shutdown zone in the fuel bundle

187

Oxide fuel rod  

International Nuclear Information System (INIS)

Spherical nuclear fuels or burnable poisons are filled into a gap between an oxide nuclear fuel pellets and a fuel can. Large grain sized pellets of high density at the center of a fuel rod transfer heat satisfactorily, to flatten the burnup degree distribution and lower the temperature of the entire fuel rod. Accordingly, the release of FP gases from the pellets is reduced. Further, the spherical fuels or the burnable poisons contribute to the control of reactivity, as well as they exhibit flowing plasticity, to prevent direct contact of the fuel can and the pellets, thereby enabling to obtain satisfactory PCMI property. (T.M.)

188

Fuel rod spacers  

International Nuclear Information System (INIS)

Purpose: To avoid the occurrence of fretting corrosions to the cladding tube of a fuel rod. Constitution: Springs are movably attached to the side walls of a fuel rod spacer. Thus, if the fuel rod vibrates or thermally expands in the turbulance of coolants, the springs themselves move following after the fuel rod and are always in contact with the cladding tube of the fuel rod. Accordingly, since no impact shocks applied to the portion in which the springs and the cladding tube of the fuel rod are abutted, occurrence of the fretting corrosions can be prevented and the fuel rod is always maintained resiliently to the proper position of the fuel spacers, it is possible to burn the fuel rod for a long period of time and increase the fuel burnup degree. (Horiuchi, T.)

189

Fuel research in Halden  

International Nuclear Information System (INIS)

The Institute for Energy Technology is an international research institute for energy and nuclear technology in Norway. This institute operates the Halden research reactor which is used for fuel research and development in all nuclear fields. This report is a short review of executed experiments of advanced fuels like inert matrix fuel, thorium fuel and Mixed OXides (MOX) fuel which have been conducted in the last three years (Authors)

190

Nuclear fuel structure and fuel behaviour  

International Nuclear Information System (INIS)

The aim of the research has been to produce information on structural properties of nuclear fuel and their effects on the fuel behaviour. The research subjects were new fuel fabrication and quality control methods, the effects of as-fabricated pellets properties on the behaviour of fuel rods, behaviour of cladding materials and irradiated cladding and structural materials. At the Technical Research Centre of Finland (VTT) the nuclear fuel structure and behaviour programme has produced data which have been utilized in procurement, behavioural analysis and surveillance of the fuel used in the Finnish nuclear power stations. In addition to our own research, data on fuel behaviour have been received by participating in the international cooperation projects, such as OECD/Halden, Studsvik-Ramp-programmes, IAEA/BEFAST II and VVER-fuel research projects. The volume of the research work financed by the Finnish Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland in the years 1987-1989 has been about 8 man years. The report is the summary report of the research work conducted in the KTM-financed nuclear fuel structure and fuel behaviour programme in the years 1987-1989

191

Fuel assembly for reactor  

International Nuclear Information System (INIS)

In a fuel assembly in which n x n fuel rods are arranged in a lattice-like configuration, assuming fuel rods at the corner on the side of control rod insertion as at (1,1) coordinate, and fuel rods on the opposite side with respect to a diagonal line as at (n,n) coordinate, fuel rods at (1,1) coordinate or (1,1), (1,2), (2,1), (1,n), (n,1) coordinates are composed of uranium fuel rods. Further, remaining fuel rods are composed of MOX fuel rods, except for burnable poison-incorporated fuel rods. That is, the number of the MOX fuel rods having lower Pu enrichment degree can be decreased based on the distribution curve for the number of MOX fuel rods and the quantity ratio of the Pu enrichment degree. In addition, since the product (quantity ratio) of the Pu enrichment degree and the number of the MOX fuel rods is small, the MOX fuel rods having lower Pu enrichment degree can be substituted with easily handlable uranium fuel rods. As a result, the kinds of Pu enrichment degree can be decreased, thereby enabling to facilitate the production and improve economical property without greatly reducing the amount of plutonium per one fuel assembly. (N.H.)

192

Fuel storage rack  

International Nuclear Information System (INIS)

If fuel enrichment degree and reactivity are increased remarkably with an aim of effective utilization of fuels in a nuclear power plant, it may become difficult to ensure subcriticality for a spent fuel storage rack now in use. It may be considered to replace the already disposed fuel storage rack with other rack of sufficient subcriticality, however, there are disadvantages that operators suffer from radiation exposure during handling of a fuel pool or fuel storage rack contaminated with radioactivity and the replacement operation needs a considerable cost. In view of the above, neutron absorbing plates are disposed detachably on the inner surface of the main body of the fuel storage rack. They can be disposed by using cranes, etc. from the outside of the fuel pool while storing spent fuels as they are. Accordingly, it is not only economical in view of continuous usage and easy operation but also the operator's exposure dose can be reduced. (N.H.)

193

Fuel storage rack  

International Nuclear Information System (INIS)

Upon inserting a fuel assembly with a large bending of a fuel channel into a square cylinder, a tapered portion at the lower part thereof is supported by being fit into a fuel seating aperture of a seating plate slidably disposed on a base. On the other hand, since the seating plate is slidable on the base, the lower end of the fuel assembly is stopped being abutted against the inner surface of the square cylinder after moving in the direction opposite to the bending of the fuel channel. That is, the center at the lower portion of the fuel assembly is displaced from the center of the square cylinder. Accordingly, the inner diameter of the fuel storage rack square cylinder can be reduced with an identical bending of the fuel assembly. This can reduce the weight and the size of the fuel storage rack. (T.M.)

194

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

195

MOX fuel assembly  

International Nuclear Information System (INIS)

The fuel assembly of the present invention comprises at least one water rod, first fuel rods filled with uranium/plutonium mixed oxide fuels, second fuel rods having axial length shorter than that of the first fuel rods and third fuel rods containing burnable poisons. If the third fuel rods are arranged on the same row and adjacent columns or on the same column and adjacent row relative to the positions where the second fuel rods are arranged or the position of the water rod replacing fuel rods, in other words, at a position extremely close to them, neutron spectrum is made softer and the neutron flux distribution is made higher. As a result, negative reactivity worth of the burnable poisons contained in the third fuel rods is enhanced, accordingly, a reactivity suppression effect comparable with that in conventional cases can be obtained by so much even if the number of the third fuel rods is reduced. The number of the MOX fuel rods is increased than a conventional case by so much as replacing the third fuel rods with the MOX fuel rods by the reduced amount thereby enabling to improve the efficiency using plutonium. (N.H.)

196

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

197

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

198

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

199

Consolidated fuel reprocessing program  

Science.gov (United States)

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

1985-04-01

200

Materials for fuel cells  

Directory of Open Access Journals (Sweden)

Full Text Available Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cells are attractive for their modular and distributed nature, and zero noise pollution. They will also play an essential role in any future hydrogen fuel economy.

Sossina M Haile

2003-03-01

 
 
 
 
201

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

202

BNFL Springfields Fuel Division  

International Nuclear Information System (INIS)

The Fuel Division of British Nuclear Fuels Ltd (BNFL) manufactures nuclear fuel elements for British Magnox and AGR power plants as well as for LWR plants. The new fuel factory - Oxide Fuel Complex (OFC), located in Springfields, is equipped with modern technology and the automation level of the factory is very high. With their quality products, BNFL aims for the new business areas. A recent example of this expansion was shown, when BNFL signed a contract to design and license new VVER-440 fuel for Finnish Loviisa and Hungarian Paks power plants. (author)

203

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

204

DUPIC fuel compatibility assessment  

Energy Technology Data Exchange (ETDEWEB)

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.

Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

2000-03-01

205

MOX fuel assembly  

International Nuclear Information System (INIS)

Since TRU has a high radioactivity, if all of the fuel rods of a fuel assembly are comprised of TRU loaded fuel rods in existent design, exposure dose is increased to makes the transportation difficult. In a fuel assembly in which MOX fuel rods are arranged in a matrix, a plurality of TRU loaded fuel rods are disposed so as to equally dividing the periphery of instrumentation guide thimble within 1st or 2nd rows around the incore instrumentation guide thimble at the central portion of a fuel assembly. Since radiations emitted from the TRU loaded fuel rods are shielded or absorbed by the peripheral MOX fuel rods, radiations of TRU released out of the fuel assembly are reduced. TRU byproduced by the operation of the reactor can be utilized effectively, and radiation emitted from the TRU loaded fuel rods can be shielded and absorbed, to reduce radioactivity of TRU released to the outside of the fuel assembly thereby greatly facilitating the handling of TRU containing-fuel assembly compared with prior cases. (N.H.)

206

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)

207

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

208

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

209

Dense fuels in Europe  

Science.gov (United States)

After a reduction of the activities on dense fuels in Europe during the late seventies, the interest in this fuel type was revived at the beginning of the eighties, when the economy of the closed FBR fuel cycle was reassessed. Strategies were developed to bring its economy into equilibrium with that of the LWR fuel cycle. The most important step was the decision to develop the European Fast Reactor (EFR) and to continue using oxide fuel at first. However, for economical reasons in an optimised fuel cycle a better fuel than the oxide should be employed. This fuel must avoid the deficiencies of the oxide but retain its advantages. Mixed nitride would satisfy this condition if it can be demonstrated that it can attain a burnup of at least 15 at%. With the know-how available in Europe it should be possible to achieve this goal with a rather limited research programme during the coming 10 years.

Blank, H.; Richter, K.; Coquerelle, M.; Matzke, Hj.; Campana, M.; Sari, C.; Ray, I. L. F.

1989-07-01

210

Fuel Air Explosives  

Directory of Open Access Journals (Sweden)

Full Text Available In this paper, important features of Fuel Air Explosives studies on different performance parameters, namely, minimum initiation energy, fuel droplet size, sensitivity to detonation etc. and current trends in this field of research have been briefly discussed.

A. Appa Rao

2014-01-01

211

Nuclear fuel rod  

International Nuclear Information System (INIS)

According to this invention, a pellet of sintered thorium dioxide is laid at the bottom of a cladding tube and sintered nuclear fuel pellets of uranium dioxide or uranium-plutonium dioxide are charged into this tube in such manner that thorium dioxide pellets are interposed between specified numbers of fuel pellets with another thorium dioxide pellet laid on the top of the stacked fuel pellets and compressed by a spring toward the fuel pellets in the tube. The tube is subsequently sealed to form a fuel element. If the fuel pellets are brought to the molten state by a thermal load, they are held by the interposed thorium dioxide pellets so as to eliminate slumping of the fuel material in the tube along the longitudinal axis during irradiation. When irradiation is terminated, the thorium dioxide pellets are transformed into a nuclear fuel. (Ohno, Y.)

212

Nuclear fuel element  

International Nuclear Information System (INIS)

Purpose: To completely prevent stress-corrosion cracking failure of a fuel can constituting a nuclear fuel element. Constitution: Super plastic material such as Zn-22Al is applied with super plastification processing by gradual cooling from a temperature higher than its eutectic point. Super plastic material layers show a high ductility under lower stresses. Accordingly, when they are bonded metallurgically to the inner surface of a fuel can, if mechanical inter actions are caused between nuclear fuel pellet and a fuel can, the stress level at the inner surface layer of the fuel can does not reach such a level as causing stress corrosion cracking. Accordingly, in the nuclear fuel element according to the present invention, no stress-corrosion cracking occurs in the fuel can, thereby enabling to obtain highly reliabile products. (Takahashi, M.)

213

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

214

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.

2012-09-11

215

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)

216

Biomass for bio fuels  

International Nuclear Information System (INIS)

You will find the justifications, potentialities and problems connected for bio fuels production. It follows a description of research activities of new generation bio fuels developed in Eni Donegani Research Centre for non conventional energies.

217

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)

218

Loviisa nuclear fuel service  

International Nuclear Information System (INIS)

The nuclear fuel service of the both units of Loviisa NPS is based on longterm fresh fuel purchasing contracts and longterm spent fuel return contracts. These contracts belong to the Soviet delivery package of Loviisa NPS and they have been made separately for the both units for their whole lifetime. The Soviet contract party is v/o Techsnabexport. Fresh fuel is ordered at the beginning of the year preceding the delivery year. The delivery takes place about one and half years earlier than the fuel is loaded into reactor. The irradiation time of the fuel is typically three years (partly two years). Spent fuel is stored at site in different storage pools five years before its returning to tbe Soviet Union. Altogether the nuclear fuel is staying at Loviisa about ten years

219

Fuel Cache at Patriot Hills  

Science.gov (United States)

... for establishment of a fuel depository of approximately sixty (60) fuel drums near the region of ... and in fact, in all uses of fuels there is the potential for fuel leakage or spillage. With ...

220

Fuel Pumping Station-Palmer  

Science.gov (United States)

... of a new fuel transfer system between Palmer's two 473,000-liter (125,000-gallon) bulk fuel storage ... pump to transfer fuel between tanks. The new pump would also be used to mix fuel which may become ...

 
 
 
 
221

Nuclear fuel cycle  

International Nuclear Information System (INIS)

Different technical and economical issues of nuclear fuel cycle are discussed, including world uranium market; uranium market price; uranium production, resources and exploration; future trends in uranium supply and demand; uranium conversion, enrichment; reactor fuel technology; control materials for water reactors; spent fuel management; prospects of plutonium use as nuclear fuel. IAEA work in this area was also highlighted. 14 refs, 10 figs, 17 tabs

222

Nuclear reactor fuel element  

International Nuclear Information System (INIS)

In order to reduce the fuel can casting one of the fuel rods has two supports, displaced relative to each other in the axial direction on the outside, between which a spacer is positioned. A lock is connected to be form-locking with the lower end of the fuel rod having the supports, which is situated transversely through the axial falling path of other fuel rods and limits axial movement in the direction of falling. (orig./HP)

223

Fuels and auxiliary materials  

International Nuclear Information System (INIS)

A brief survey is given of the problems of fuels, fuel cans, absorption and moderator materials proceeding from the papers presented at the 1971 4th Geneva Conference on the Peaceful Uses of Nuclear Energy and the 1970 IAEA Conference in New York. Attention is focused on the behaviour of fuel and fuel can materials for thermal and fast reactors during irradiation, radiation stability of absorption materials and the effects of radiation on concrete and on moderator materials. (Z.M.)

224

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

225

Fused carbonate fuel cell  

International Nuclear Information System (INIS)

A fused carbonate type of fuel cell comprising an electrolytic body retaining an electrolyte therein which is arranged between an anode and a cathode, where electricity is electrochemically generated by feeding fuel gas and an oxidant to a fuel chamber arranged on the anode side and an oxidant chamber arranged on the cathode side, respectively, said fuel cell being characterized in that the electrolytic body comprises an electrolyte, an electrolyte-holding member for holding the electrolyte and an inorganic binder

226

Fuels for research reactors  

International Nuclear Information System (INIS)

The research reactors are indispensable in several scientific studies. They make use of fuels based on highly enriched uranium and aluminium. The CERCA Company, owned by Framatome (51%) and Cogema (49%), holds a predominant place on the fuel market. The paper presents technical generalities, the industrial configuration in the production of highly enriched fuels for research reactors and a table containing the principal clients, the reactor types and the enrichment level of the fuels supplied by CERCA Company

227

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)

228

Metallic fuel development  

International Nuclear Information System (INIS)

Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

229

The PWR fuel cycle  

International Nuclear Information System (INIS)

This symposium on the fuel cycle of pressurized water reactors (PWR) deals in turn with the extraction and isotopic separation of the uranium, the fabrication and testing of the fuels, the use made of such fuels, the loading, unloading and reprocessing and, lastly, the radioactive waste

230

Gelled fuel simulant  

International Nuclear Information System (INIS)

A relatively stable inert simulant formulation for a hazardous metallized fuel has the density, shear rate and yield stress of the duplicated fuel. This formulation provides inexpensive and safe testing of exploratory hydraulic studies, or testing of the mechanical strength of containers, plumbing, etc., in which the metallized fuels are to be used

231

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.

2012-07-20

232

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.

2012-09-12

233

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)

234

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.

2012-01-03

235

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.

Tyson, Peter; Novasciencenow

236

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

237

CANDU fuel performance  

International Nuclear Information System (INIS)

The paper presents a review of CANDU fuel performance including a 28-element bundle for Pickering reactors, a 37-element bundle for the Bruce and Darlington reactors, and a 37-element bundle for the CANDU-6 reactors. Special emphasis is given to the analysis of fuel defect formation and propagation and definition of fuel element operating thresholds for normal operation and accident conditions. (author)

238

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.)

239

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.

240

Neutronic fuel element fabrication  

Science.gov (United States)

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

Korton, George (Cincinnati, OH)

2004-02-24

 
 
 
 
241

Reactor fuel element and fuel assembly  

International Nuclear Information System (INIS)

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

242

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.

243

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

244

Transportation of nuclear fuel  

International Nuclear Information System (INIS)

Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

245

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

246

Low void reactivity fuel  

International Nuclear Information System (INIS)

Low void reactivity fuel (LVRF) is an important advanced fuel development. It is currently being qualified for introduction into Bruce Power's Unit B reactors, and it forms the basis for the Advanced CANDU Reactor (ACR) fuel design. This paper describes the LVRF concept, the variants that AECL considered before adopting the current reference design, the generic qualification undertaken to establish the technical feasibility of this new fuel type, applications, and the ongoing advanced fuel development in AECL that will support enhancements of this product in the future. (author)

247

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

248

Thorium fuel cycle radiotoxicity  

International Nuclear Information System (INIS)

The aim of this study was to find radiotoxicity, relative radiotoxicity index and index of the residual risk of thorium fuel cycle facilities. The calculation was made for two extreme cycles. Extreme, therefore, that the first fuel cycle is not considered a shutdown during irradiation of fuel in the reactor core. Conversely, in the second fuel cycle time between shutdowns was extremely long. To investigate and decide if shutdown during irradiation of fuel in the reactor will have an impact on the development and progress relative radiotoxicity and residual risk. (author)

249

Automatic fuel number reader  

International Nuclear Information System (INIS)

Optical and ultrasonic fuel number readers have been developed to realize the efficient and automatic confirmation and verification of fuel numbers, thereby to reduce mental load and radiation exposure of operators who engage in the confirmation and verification task. This task is carried out as a part of the safeguards in spent fuel storage facilities. The optical fuel number imaging device has been designed to illuminate a fuel number which is carved on the top side surface of each PWR fuel assembly and to visualize it with an underwater TV camera. In this research, the prototype periscope of an optical device using a side view mirror has been developed to achieve high speed verification of fuel numbers. An ultrasonic fuel number imaging device has been developed for a supplement of the optical method, when fuel numbers can be hardly recognized due to the deposition of crud and others. The image of a fuel number can be obtained by scanning focused ultrasonic wave on the surface of the fuel number part. In this research, a one-dimensional array type piezoelectric sensor has been developed, which enabled the electronic scanning of focused ultrasonic wave. (K.I.)

250

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

251

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.)

252

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

253

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.)

254

Failed fuel element detector  

International Nuclear Information System (INIS)

In the failed fuel element detector of this invention, a truck capable of moving on a bridge across a coolant water tank of a pool type reactor has a pipe extending downward to or near the coolant drain port of an arbitrarily selected fuel element. Coolant in the element is drawn by a pump through this pipe and collected in a reservoir located in the truck. The radioactivity of the coolant in the reservoir is measured to determine the content of any radioactive effluent elected from the fuel element, into the coolant. Thus, a failed fuel element can be located using this detector system without extracting fuel elements from the reactor core. Labor required in the locating of failed elements. This detector in the fuel assembly may thus be reduced system may be positioned so as to monitor the radioactivity of the coolant in the tank, sudden increases in radioactivity indicate fuel element failure. (JPN)

255

Reference thorium fuel cycle  

International Nuclear Information System (INIS)

The purpose of this report is to define a preliminary thorium fuel cycle to serve as a common basis for beginning development work on October 1, 1977, at participating ERDA laboratories, universities, and commercial facilities. Characteristics of the reference fuel cycle for the Thorium Fuel Cycle Technology (TFCT) program are: fissile uranium will be denatured by mixing with 238U; chemical processing plant design will be based on the assumption that plants are located in secure areas; plutonium will be recycled within these secure areas; thorium will be recycled with recovered uranium and plutonium; the head end of the chemical processing plant will handle a variety of core and blanket fuel assembly designs for light water reactors and heavy water reactors; the fuel form will be a homogeneous mixture of uranium and thorium oxide powders pressed into pellets; fuel cladding will be Zircaloy; and MgO will be added to the fuel to improve the thorium dissolving characteristics

256

Nuclear fuel vibrocompacting  

International Nuclear Information System (INIS)

Prospects of using vibrocompacting in production of fuel elements with mixed oxide fuel are discussed. The role of homogeneity of granulometric composition and form compacted powder particle is considered. Irradiation of a large number of vibrocompacted fuel elements in the BOR-60 reactor has not shown changes in their performances. Technology of fuel vibrocompacting is concluded to be used most likely in production of fuel elements with regenerated fuel for fast breeder reacotors. The described technology complies with reguirements of rational, reliable and remote - controlled production. Industrial use of the considered method depends mainly on the possibility of preparing well compacted powder of homogeneous mixed oxide fuel with stable reproducible characteristics. Special attention should be paid to assurance of high density and constancy of dimensions for certain fractions of spherical particles

257

BN-600 fuel elements and fuel assemblies operating experience  

International Nuclear Information System (INIS)

Consideration is given to the data on fuel burnup of standard fuel assemblies of the BN-600 reactor first core charge and that for modified core; data on operation ability of fuel assemblies of the first charge type are given. Data on main results of primary post-irradiation examination of fuel assemblies and fuel elements and maximal values of fuel burnup, achieved in particular fuel assemblies of BN-600 reactor are presented. 4 figs.; 1 tab

258

Method of decladding spent fuel  

International Nuclear Information System (INIS)

Purpose: To enable to safety and easy decladding of nuclear fuels thereby reduce the processing cost. Constitution: Upon dismantling of a spent fuel rod, the fuel rod is heated at least to such a temperature that the ductility of a fuel can is recovered, then transported by using seizing rollers, by which the fuel rod is pressurized from the outer circumference to break the nuclear fuels at the inside thereof. Then, the destructed fuels are recovered from both ends of the fuel can. With such a constitution, since the ductility of the fuel can is recovered by heating, when the fuel rod is passed through the rollers in this state, the fuel can is deformed to destroy the nuclear fuels at the inside thereof. Since the nuclear fuels are destroyed into small pieces, they can be taken out easily from both ends of the fuel can. (Kawakami, Y.)

259

Fuel safety research 1999  

International Nuclear Information System (INIS)

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)

260

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

 
 
 
 
261

Diesel fuel filtration system  

International Nuclear Information System (INIS)

The American nuclear utility industry is subject to tight regulations on the quality of diesel fuel that is stored at nuclear generating stations. This fuel is required to supply safety-related emergency diesel generators--the backup power systems associated with the safe shutdown of reactors. One important parameter being regulated is the level of particulate contamination in the diesel fuel. Carbon particulate is a natural byproduct of aging diesel fuel. Carbon particulate precipitates from the fuel's hydrocarbons, then remains suspended or settles to the bottom of fuel oil storage tanks. If the carbon particulate is not removed, unacceptable levels of particulate contamination will eventually occur. The oil must be discarded or filtered. Having an outside contractor come to the plant to filter the diesel fuel can be costly and time consuming. Time is an even more critical factor if a nuclear plant is in a Limiting Condition of Operation (LCO) situation. A most effective way to reduce both cost and risk is for a utility to build and install its own diesel fuel filtration system. The cost savings associated with designing, fabricating and operating the system inhouse can be significant, and the value of reducing the risk of reactor shutdown because of uncertified diesel fuel may be even higher. This article describes such a fuel filtering system

262

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.)

263

Fuel loading method  

International Nuclear Information System (INIS)

Among initially loaded fuel assemblies, fuel assemblies having an average enrichment degree higher than that of exchange fuels are taken out from the reactor core after combustion for 3 cycles, and then loaded again to the reactor core after passing one or more cycles. Since the high enrichment degree fuels taken out after three cycles have higher enrichment degree and have shorter combustion period and lower burnup degree compared with the exchange fuels which are burnt for four to five cycles, most of fission products are remained. Accordingly, if the high enrichment degree fuels are loaded again after passing one or more cycles, the take-up burnup degree of the initially loaded fuels can be sufficiently increased, as well as the number of exchange fuels can be reduced. In addition, since the burnup degree is increased as the cycle is proceeded, the difference of the burnup degree between the high enrichment degree fuels to be loaded again and the enrichment degree fuels taken up instead of them is increased. (N.H.)

264

Fuel related risks; Braenslerisker  

Energy Technology Data Exchange (ETDEWEB)

The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from the interviews was supplemented with examples from the literature

Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

2012-02-15

265

Manufacture of nuclear fuel compacts  

International Nuclear Information System (INIS)

Nuclear fuel rods are manufactured utilizing a graphite flourpitch matrix formulation containing an additive. The matrix formulation has a decreased viscosity at fabrication temperatures which permits manufacture of the fuel rods with lower fabrication pressures. Also, the matrix formulation does not cause the fuel rod to adhere or bond to the fuel element during heat treatment of the fuel rod in the fuel element. The nuclear fuel rods are suitable for use in high temperature gas cooled nuclear

266

An intelligent spent fuel database for BWR fuels  

International Nuclear Information System (INIS)

The present aim is to establish an intelligent database of Spent Fuel Data (including physical fuel data and reactor operating history information) to support burnup credit analyses for Boiling Water Reactor Fuel. At a later date, information of Pressurized Water Reactor Fuel and existing Post-Irradiation Examination (PIE) data for benchmarking fuel composition calculations may be integrated into the database. (author)

267

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)

268

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

269

Fuel cell generator with fuel electrodes that control on-cell fuel reformation  

Science.gov (United States)

A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

Ruka, Roswell J. (Pittsburgh, PA); Basel, Richard A. (Pittsburgh, PA); Zhang, Gong (Murrysville, PA)

2011-10-25

270

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.)

271

Nuclear fuel rod  

International Nuclear Information System (INIS)

Fuel pellets formed by incorporating transuranium elements in metal uranium are disposed above oxide uranium fuel pellets. The transuranium-incorporating metal uranium fuel pellets are formed by mixing at least one of Np-237, Am-241 and Am-243 to the metal uranium not absorbing high energy neutrons by oxygen. For example, in a BWR type reactor, neutron spectrum is hard at the upper portion of the nuclear fuel rod. Accordingly, the transuranium-incorporating metal uranium fuel pellets are charged at the upper portion of the nuclear fuel rod, having high energy neutrons, in other words, having hard neutron spectrum. With such a constitution, transuranium elements such as Np and Am can be annihilated effectively. (I.N.)

272

Production of liquid fuels  

International Nuclear Information System (INIS)

The greenhouse gas emissions associated with the production and usage of a number of liquid fuels are reviewed, including petroleum-based transport fuels and synthetic fuels derived from natural gas, coal and shale. Comments are also offered on the role of biomass-derived fuels. The methodology employed is similar to that used by DeLuchi and Sperling (1988), but extended to include the production of hydrocarbon fuels from coal and shale. While most of their conclusions are supported, some differences are noted. Most of the current interest in modifying the usage and formulation of transport fuels, stems from the benefits claimed to be achievable for local air quality rather than from reduced greenhouse gas emissions. The changes proposed, could add to the greenhouse burden while others may diminish it. 26 refs., 3 tabs., 2 figs

273

Nuclear fuel assembly  

International Nuclear Information System (INIS)

In FBR type reactors, fuel boundle-wrapper tube interaction (BDI) is caused, particularly, near the axial center of the reactor core due to irradiation swelling and creeping. Accordingly, in the present invention, a portion of fuel rods are constituted such that depleted uranium pellets are disposed near the axial center of the reactor core. Further, fuel rods not disposed with depleted uranium pellets near the axial center of the reactor core are disposed such that they are not in adjacent with each other at least in the radial peripheral portion. Thus, even if fuel rods are brought into contact each other by BDI, abrupt temperature rise of fuel cans can be moderated. Further, occurrence of the fuel can failure can remarkably be reduced, to improve the safety. (T.M.)

274

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.)

275

CANDU fuel cycle flexibility  

International Nuclear Information System (INIS)

High neutron economy, on-power refuelling, and a simple bundle design provide a high degree of flexibility that enable CANDU (CANada Deuterium Uranium; registered trademark) reactors to be fuelled with a wide variety of fuel types. Near-term applications include the use of slightly enriched uranium and recovered uranium from reprocessed spent Light Water Reactor (LWR) fuel. Plutonium and other actinides arising from various sources, including spent LWR fuel, can be accommodated, and weapons-origin plutonium could be destroyed by burning in CANDU. In the DUPIC fuel cycle, a dry processing method would convert spent Pressurized Water Reactor fuel to CANDU fuel. The thorium cycle remains of strategic interest in CANDU to ensure long-term resource availability, and would be of specific interest to those countries possessing large thorium reserves, but limited uranium resources. 21 refs

276

NOVA: Fuel Cells  

Science.gov (United States)

This 14-minute video explores the technology of the hydrogen fuel cell designed for use in automobiles. It delves into the promise of the technology, hurdles (such as building filling stations), and challenges of finding green ways to produce hydrogen. You'll also find an animated tutorial on how fuel cells work, a clickable schematic view of a fuel cell installed in a car, and an a question/answer set featuring an expert in energy conversion devices.

2014-03-30

277

Fuel element transport container  

International Nuclear Information System (INIS)

The transport system is employed in a fuel cycle center for storage, relocation or reprocessing of LWR fuel elements. It consists of a mast movable in the vertical direction, capable of horizontal displacement and carrying a swivel boom at its bottom end. This boom is used to move fuel elements from one pool into the other. The mast is braced against lateral forces in a horizontal rail. (DG)

278

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.)

279

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

280

Vented nuclear fuel element  

International Nuclear Information System (INIS)

A nuclear fuel cell is described for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel

 
 
 
 
281

Nuclear reactor fuel elements  

International Nuclear Information System (INIS)

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

282

Nuclear fuel accounting  

International Nuclear Information System (INIS)

After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW)

283

Liquid fuel cells.  

Science.gov (United States)

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. PMID:25247123

Soloveichik, Grigorii L

2014-01-01

284

European fuel development trends  

International Nuclear Information System (INIS)

ABB has looked carefully at what type of improvements utilities want to see during the next ten years: fuel reliability - zero defect goal and trouble free fuel utilization; economic utilization - driven by high backend costs; operating flexibility - load follow and cycle length changes; safe operation -thorough verification process; plant improvements - longer cycles, power uprating. The improved fuel designs have been made more forgiving to accommodate anticipated more demanding operating conditions and to increase flexibility in fuel cycle design. The overall development trend is mainly dictated by the very high backend costs, leading to ever higher discharge burnup. (orig./HP)

285

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)

286

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.)

287

ABB high burnup fuel  

International Nuclear Information System (INIS)

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with GuardianTM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

288

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)

289

Fuel assembly reconstitution  

Energy Technology Data Exchange (ETDEWEB)

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)

Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos, E-mail: mongeor@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil)

2009-07-01

290

Nuclear reactor fueling system  

International Nuclear Information System (INIS)

Nuclear fuel transfer gear to move nuclear fuel assemblies between a storage vessel and a nuclear reactor vessel placed against this storage vessel, each of these vessels having coolant inlets and outputs for the coolant to flow through them. This gear includes: - a connexion system forming a passage between the reactor vessel and the storage vessel located below the level of the coolant, so as to permit the transfer of the fuel assemblies between the reactor vessel and the storage vessel whilst maintaining these fuel assemblies completely immersed in the coolant; - a first system for the transfer of fuel assemblies fitted in the reactor vessel above the reactor core; - and a second fuel assembly transfer system fitted in the storage vessel above the fuel assembly storage area. These facilities are characterised in that the first and second fuel assembly transfer systems include a first and second respective revolving appliance, to hold the fuel assemblies and to align them with the connexion system so as to transfer them through the passage formed by this connexion system

291

Reactor fuel element  

International Nuclear Information System (INIS)

Purpose: To reduce the pressure loss in the reactor core and improve the heat-removing performance in fuel elements for use in pebble bed type high temperature gas reactors. Constitution: The fuel element according to the present invention is prepared by molding cladded fuel particles being dispersed in a graphite matrix and then sintering them while forming therearound a shell made of heat resistance material (graphite or ceramics) fabricated into porous structure. Since the shell in the fuel element is porous, cooling gas permeability is improved in a case where the spheres are stacked in the reactor core to reduce the pressure loss in the reactor core portion. (Kamimura, M.)

292

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

293

Reference thorium fuel cycle  

International Nuclear Information System (INIS)

In the reference fuel cycle for the TFCT program, fissile U will be denatured by mixing with 238U; the plants will be located in secure areas, with Pu being recycled within these secure areas; Th will be recycled with recovered U and Pu; the head end will handle a variety of core and blanket fuel assembly designs for LWRs and HWRs; the fuel may be a homogeneous mixture either of U and Th oxide pellets or sol-gel microspheres; the cladding will be Zircaloy; and MgO may be added to the fuel to improve Th dissolution. Th is being considered as the fertile component of fuel in order to increase proliferation resistance. Spent U recovered from Th-based fuels must be re-enriched before recycle to prevent very rapid buildup of 238U. Stainless steel will be considered as a backup to Zircaloy cladding in case Zr is incompatible with commercial aqueous dissolution. Storage of recovered irradiated Th will be considered as a backup to its use in the recycle of recovered Pu and U. Estimates are made of the time for introducing the Th fuel cycle into the LWR power industry. Since U fuel exposures in LWRs are likely to increase from 30,000 to 50,000 MWD/MT, the Th reprocessing plant should also be designed for Th fuel with 50,000 MWD/MT exposure

294

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

295

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)

296

Spent fuel storage chamber  

International Nuclear Information System (INIS)

In a dry spent nuclear fuel storage chamber, an atmosphere in a closed loop comprising storage cell/heated air collecting chamber/cooling air circulation path is filled with gases having a high thermal radiation absorbing performance. Heat released from the spent fuels heats a cylindrical vessel, gases in contact with the peripheral surface thereof and metal blocks constituting the storage cell. Since the gases having highly heat absorbing performance are filled, they are heated by absorbing radiation heat of the spent fuels, to improve the heat dissipation efficiency of the spent fuels. Accordingly, even if the heat generation amount of the spent fuels is great, the temperature elevation can be suppressed since the heat dissipation efficiency of the spent fuels is great due to radiation absorption. In addition, a phenomenon that the temperature of the cylindrical vessel is raised can be suppressed. As a result, fuels or mixed oxide fuels of a high burnup degree having greater heat generation amount compared with usual fuels can be stored safely and economically. (N.H.)

297

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

298

Vent type fuel pin and fuel assembly  

International Nuclear Information System (INIS)

An end plug of each of fuel pins is provided with a property of allowing only gases to permeate, and a temperature sensitive valve mechanism which closes at high temperature and opens at low temperature is disposed between the end plug and a gas plenum. Further, in a fuel assembly, a common pas plenum is disposed at the inside of a wrapper tube and each of the fuel pins is closed at one end by the end plug while being opened at the other end so as to be in communication with a common gas plenum. Then, a member allowing only gases to permeate is disposed to the exit of the common gas plenum, and a temperature sensitive valve mechanism which closes at high temperature and open at low temperature is disposed. This inhibits radioactive gases from emitting during reaction operation (high temperature state) and does not hinder the detection of failed fuels although it is a vent type structure. Further, since accumulated radioactive gases are released during shutdown of the reactor (low temperature state), increase of the inner pressure in the fuel pin can be reduced to about 1/3 of that of the existent sealed type pin. (T.M.)

299

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.)

300

Numerical simulation of fuel assemblies containing pebble fuel elements  

Science.gov (United States)

A numerical calculation of the new fuel assembly for a VVER-1000 reactor containing fuel on the basis of pebble fuel elements is carried out. The reactor core pressure drop coefficients in the axial and radial directions are studied.

Bocharova, E. V.; Tokarev, Yu. N.; Komov, A. T.

2010-12-01

 
 
 
 
301

Fuel Cells Vehicle Systems Analysis (Fuel Cell Freeze Investigation)  

Energy Technology Data Exchange (ETDEWEB)

Presentation on Fuel Cells Vehicle Systems Analysis (Fuel Cell Freeze Investigation) for the 2005 Hydrogen, Fuel Cells & Infrastructure Technologies Program Annual Review held in Arlington, Virginia on May 23-26, 2005.

Pesaran, A.; Kim, G.; Markel, T.; Wipke, K.

2005-05-01

302

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)

303

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

304

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

305

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

306

Fuel safety research 2000  

International Nuclear Information System (INIS)

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)

307

Nondestructive measurements on spent fuel for the nuclear fuel cycle  

International Nuclear Information System (INIS)

Nondestructive measurements on spent fuel are being developed to meet safeguards and materials managment requirements at nuclear facilities. Spent-fuel measurement technology and its applications are reviewed

308

Fuel cycle economical improvement by reaching high fuel burnup  

International Nuclear Information System (INIS)

Improvements of fuel utilization in the light water reactors, burnup increase have led to a necessity to revise strategic approaches of the fuel cycle development. Different trends of the fuel cycle development are necessary to consider in accordance with the type of reactors used, the uranium market and other features that correspond to the nuclear and economic aspects of the fuel cycle. The fuel burnup step-by-step extension Program that successfully are being realized by the leading, firms - fuel manufacturers and the research centres allow to say that there are no serious technical obstacles for licensing in the near future of water cooling reactors fuel rod burnup (average) limit to 65-70 MWd/kgU and fuel assembly (average) limit to (60-65) MWd/kgU. The operating experience of Ukrainian NPPs with WWER-1000 is 130 reactor * years. At the beginning of 1999, a total quantity of the fuel FA discharged during all time of operation of 11 reactors was 5819 (110 fuel cycles). Economical improvement is reached by increase of fuel burn-up by using of some FA of 3 fuel cycles design in 4th fuel loading cycle. Fuel reliability is satisfactory. The further improvement of FA is necessary, that will allow to reduce the front-end fuel cycle cost (specific natural uranium expenditure), to reduce spent fuel amount and, respectively, the fuel cycle back end costs, and to increase burn-up of the fuel. (author)

309

Fuel spill reports-Format  

Science.gov (United States)

Title : Fuel spill reports-Format Type : Antarctic EAM NSF Org: OD / OPP Date : February 05, 1991 ... Action Memorandum (Threshold for Fuel Spill Reports/Format for Fuel Spill Reports) To: Files (S.7 - ...

310

Fuel cells: Problems and prospects  

Digital Repository Infrastructure Vision for European Research (DRIVER)

n recent years, fuel cell technology has advanced significantly. Field trials on certain types of fuel cells have shown promise for electrical use. This article reviews the electrochemistry, problems and prospects of fuel cell systems.

Shukla, Ak; Ramesh, Kv; Kannan, Am

1986-01-01

311

MOX fuel assembly  

International Nuclear Information System (INIS)

MOX fuel rods containing fissionable Pu are bundled in substantially in a normal square matrix except for four corners, in which MOX fuel rods at higher enrichment degree are disposed in the central portion than those in the outer circumferential portion. Namely, fuel rods at medium enrichment degree are disposed at the outer circumferential portion except for the four corners and to the outer side of control rod guide thimble at the outer circumference. In addition, fuel rods at high enrichment degree are disposed at the central portion. Burnable poison (BP) rods are disposed at each of the corners. The BP rods may be disposed to at least four corners and, in addition, at the outer circumference in adjacent with the corners instead of fuel rods. With such a constitution, the kind of the MOX fuel rods used in an LWR type reactor can be decreased to two kinds without substantially reducing the enrichment degree of the fissionable Pu as a fuel assembly, thereby enabling to reduce the fuel molding and fabrication cost. (I.N.)

312

Stabilized fuel slurry  

Energy Technology Data Exchange (ETDEWEB)

A fuel slurry comprising a mixture of a fuel oil and pulverized coal may be effectively stabilized with a small amount of adducts of alkylene oxide and an alcohol, an amine, a carboxylic acid or a phenol, or inorganic acid esters of said adducts or crosslinked products of said adducts or said inorganic acid esters.

Aoike, K.; Honjo, S.; Naka, A.

1981-02-17

313

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.)

314

Spent fuel storage rack  

International Nuclear Information System (INIS)

The present invention provides a storage rack capable of improving fuel storage density of spent fuel assemblies generated in a BWR type power plant. In the spent storage rack, a plurality of storage cells for containing fuel assemblies one by one are constituted by shielding members, the shielding member is made of stainless steel containing 1wt% of more of boron, and the shielding member is determined up to nominal 5mm thickness. An austenite stainless steel plate to which boron is added at a high concentration acts so as to make the arrangement of the fuel rack cells regular square lattice have the geometrically highest density, and acts so as to reduce the distance between fuel rack cell and the stored fuels by absorbing thermal neutrons generated from the spent fuels. The plate having up to nominal 5mm thickness makes the voids present in the material finer during ordinary rolling and acts so as to provide mechanical properties required as a fuel rack. (I.S.)

315

Nuclear fuel can  

International Nuclear Information System (INIS)

Purpose: To provide nuclear fuel cans capable of load following operation with no worry of stress corrosion cracking. Constitution: If the reactor power rises abruptly in a stage where fuel burnup degree is increased, chemical reactions are taken place between the fuel can and the corrosive nuclear fission products and, at the same time, thermal stresses are exerted on the fuel can due to the heat expansion of nuclear fuel pellets, by which stress corrosion crackings are caused by the combined effects. In view of the above, the inner surface of the fuel can substrate made of zirconium alloy is lined by depositing a graphite layer of 0.5 - 100 ?m thickness. Since the graphite lining can prevent the contact between the fuel can substrate and the corrosive nuclear fission products and the lubricating property of the graphite lining can reduce the frictional coefficient upon contact due to the heat expansion, local stresses can be reduced. Thus, the reliability of the nuclear fuel elements can be improved to practice the load following operation. (Kamimura, M.)

316

Nuclear fuel transport flask  

International Nuclear Information System (INIS)

A nuclear fuel transport flask has a fuel containing compartment which is supplied with decontaminating fluid via inlet passageways and tubes which discharge into the compartment. The outlet from the compartment is via a box and outlet passageways within end cap. The passageways are conveniently situated at the same end of the flask. (author)

317

Bottom mounted fuel holddown  

International Nuclear Information System (INIS)

The present invention relates to nuclear reactor fuel assemblies, and in particular to an apparatus for holding a fuel assembly down against a core support stand. The locking can be selectively controlled from the upper end of the assembly without producing excessive wear on the support stand structure. In the event of failure of a locking component, unlocking can be accomplished remotely

318

Fuel sorting evaluation  

Energy Technology Data Exchange (ETDEWEB)

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.

Pajunen, A.L.

1996-03-12

319

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

320

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.

2012-04-27

 
 
 
 
321

Magnetic fuel line device  

Energy Technology Data Exchange (ETDEWEB)

A device is described which is positionable adjacent a fuel line of a fuel consuming apparatus for acting on fuel flowing therein for increasing the fuel efficiency of the apparatus and for reducing pollution emissions therefrom. The device consists of: a housing comprising a body of non-magnetic material, the housing defining a longitudinally extending passage for receiving the fuel line; means for securing the housing to the fuel line; a magnet formed from a magnetic material magnetized with one pole on one longitudinal face thereof and the other pole on the opposite longitudinal face thereof. The magnet is embedded in the body of non-magnetic material in the housing with one of the poles adjacent and parallel to the longitudinally extending passage for positioning the one of the poles adjacent and parallel to one side of the fuel line when the device is secured thereto. Theere are no other magnets disposed about the fuel line between the end faces of the magnet.

Mitchell, J.; Ament, C.E.

1986-02-25

322

Nuclear fuel element  

International Nuclear Information System (INIS)

A nuclear fuel pin is formed of an elongated metallic tube, closed at each end, containing stacked fuel pellets and material including little, or no, fissionable material, and formed into porous, or bubbled, microspheres placed in the annulus between the pellets and the internal wall of the tube

323

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

324

Fuel particle coating data  

International Nuclear Information System (INIS)

Development of coating on nuclear fuel particles for the High-Temperature Fuels Technology program at the Los Alamos Scientific Laboratory included process studies for low-density porous and high-density isotropic carbon coats, and for ZrC and ''alloy'' C/ZrC coats. This report documents the data generated by these studies

325

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)

326

International fuel bank  

International Nuclear Information System (INIS)

The working group discusses the establishment of an international bank for nuclear fuels. The statements by representatives of seven countries discuss the specific features of a bank of this kind which is set up to facilitate access to nuclear fuels but also to permit a more rigid control in the sense of the non-proliferation philosophy

327

Renewable fuels. Proceedings  

International Nuclear Information System (INIS)

Renewable fuels are an interesting alternative to conventional agriculture. Fuels from flax, rapeseed, salix and miscanthus will not only help save fossil resources but are also on economically acceptable may of using agricultural land that would otherwise be left barren. (orig.)

328

Fuel pin plenum spring  

International Nuclear Information System (INIS)

A fuel pin including plenum spring consisting of a material having a temperature dependent spring constant so selected as to substantially reduce the spring force when the spring is at reactor operating temperature is described. With this arrangement, the spring force applied during shipping may be relatively high without overstressing the fuel pellets during reactor operation. (author)

329

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

330

Fuel assembly for reactor  

International Nuclear Information System (INIS)

In a Gd-added fuel assembly, the axial length is divided into a plurality of regions in which the upper region comprises Pu-enriched MOX + Gd fuels and the lower region comprises UO2 + Gd fuels using enriched uranium. Infinite multiplication factor K? in the lower region is increased uniformly compared with a case of using a MOX + Gd fuel assembly axially uniformly when burning is proceeded, and axial thermal power distribution is expanded downwardly. On the other hand, conversion of U-238 to Pu-239 along with burning is generally greater in the upper region of a fuel assembly having high coolants void coefficient, and the axial thermal power distribution is expanded upwardly along with burning. The axial thermal power distribution is made substantially uniform with no upper expansion throughout burning due to these two conflicting effects. (T.M.)

331

Assessment of automotive fuels  

International Nuclear Information System (INIS)

Energy demand all over the world increases steadily and, within the next decades, is almost completely met by fossil fuels. This poses increasing pressure on oil supply and reserves. Concomitant is the concern about environmental pollution, especially by carbon dioxide from fossil fuel combustion, with the risk of global warming. Environmental well-being requires a modified mix of energy sources to emit less carbon dioxide, starting with a move to natural gas and ending with the market penetration of renewable energies. Efforts should focus on advanced oil and gas production and processing technologies and on regeneratively produced fuels like hydrogen or bio-fuels as well. Within the framework of an industrial initiative in Germany, a process of defining one or two alternative fuels was started, to bring them into the market within the next years. (orig.)

332

Extended fuel cycle length  

International Nuclear Information System (INIS)

Extended fuel cycle length and burnup are currently offered by Framatome and Fragema in order to satisfy the needs of the utilities in terms of fuel cycle cost and of overall systems cost optimization. We intend to point out the consequences of an increased fuel cycle length and burnup on reactor safety, in order to determine whether the bounding safety analyses presented in the Safety Analysis Report are applicable and to evaluate the effect on plant licensing. This paper presents the results of this examination. The first part indicates the consequences of increased fuel cycle length and burnup on the nuclear data used in the bounding accident analyses. In the second part of this paper, the required safety reanalyses are presented and the impact on the safety margins of different fuel management strategies is examined. In addition, systems modifications which can be required are indicated

333

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

334

The future fuel cycle  

International Nuclear Information System (INIS)

Sustainability is a major driver of France nuclear strategy, among which fuel cycle policy plays a major role. France nuclear fuel cycle strategy is based on the recycling of spent fuel in order to save energy resources and to offer a better management of nuclear wastes. This relies on decades' feedback experience. Closing the uranium plutonium fuel cycle, in a recurrent way, in fast neutron reactors is one of the keys to sustainable nuclear energy, by preventing the accumulation of sensitive materials, preserving natural resources - by more than a factor 100 - and limiting the ultimate waste volume. Management options for the future fuel cycle are also being investigated in an approach consistent with studies for the Astrid 4. generation sodium cooled fast reactor project. The goal is to prepare all the options for the management of nuclear materials in the fleets of fast reactors, by developing the most advanced processes. (author)

335

Alternative fuel transit buses  

Energy Technology Data Exchange (ETDEWEB)

The National Renewable Energy Laboratory (NREL) is a U.S. Department of Energy (DOE) national laboratory; this project was funded by DOE. One of NREL`s missions is to objectively evaluate the performance, emissions, and operating costs of alternative fuel vehicles so fleet managers can make informed decisions when purchasing them. Alternative fuels have made greater inroads into the transit bus market than into any other. Each year, the American Public Transit Association (APTA) surveys its members on their inventory and buying plans. The latest APTA data show that about 4% of the 50,000 transit buses in its survey run on an alternative fuel. Furthermore, 1 in 5 of the new transit buses that members have on order are alternative fuel buses. This program was designed to comprehensively and objectively evaluate the alternative fuels in use in the industry.

Motta, R.; Norton, P.; Kelly, K. [and others

1996-10-01

336

Fuel decontaminating device  

International Nuclear Information System (INIS)

Purpose: To remove metal oxides deposited on the surface of fuel rod by using reducing agent without dismantling a fuel assembly. Constitution: A decontaminating liquid supply tool and a liquid removal tool are mounted to one and the other ends of a metal cylinder containing a nuclear fuel assembly in order to flow the decontaminating liquid for chemically dissolving metal oxides. The metal oxides on the surface of the fuels are removed while being dissolved in a chemical reduction manner. In order to increase the dissolving power of the decontaminating liquid, electrolytes reduced by the cathodic electrolysis are used. In this method, decontamination can be carried out even to the gaps or the portions behind the obstacles which can not be cleaned by the mechanical shocks in conventional mechanical elimination method. Further, it is no more necessary to dismantle the fuel assembly thus causing no surrounding contamination. Further, the device is simple and the worker's exposure can be decreased. (Moriyama, K.)

337

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

338

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO2 fuel substances. Constitution: Two or more additives selected from Al2O3, B2O, CaO, MgO, SiO2, Na2O and P2O5 are added by the total weight of 2 - 5% to fuel substances consisting of UO2 or a mixture of UO2 and PuO2. When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

339

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To decrease the strain due to fuel-cladding interactions by adapting to attain certain relationships among the diameter/height ratio in a fuel pellet, the maximum dimension of a diagonal line passing through the pellet center, diameter of the fuel pellet, gap and the like. Constitution: Fuel pellets filled in a fuel cladding tube are formed by adding and compounding polyvinyl alcohol or the like with fine powder of uranium dioxide, pelletized and compression molded into a predetermined shape and then sintered after degrease under a hydrogen stream. The sintered pellets are cylindrical and satisfy the following relations: H/D < 0.7, where D is a diameter (mm) and H is a height (mm), and l < (D + G), where l is the maximum dimension (mm) of a diagonal line passing through the pellet center, and G is the diametrical gold gap (mm) between the cladding tube and the pellet. (Moriyama, K.)

340

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

 
 
 
 
341

FUEL CELLS IN ENERGY PRODUCTION  

Digital Repository Infrastructure Vision for European Research (DRIVER)

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

342

Biological and microbial fuel cells  

Digital Repository Infrastructure Vision for European Research (DRIVER)

Biological fuel cells have attracted increasing interest in recent years because of their applications in environmental treatment, energy recovery, and small-scale power sources. Biological fuel cells are capable of producing electricity in the same way as a chemical fuel cell: there is a constant supply of fuel into the anode and a constant supply of oxidant into the cathode; however, typically the fuel is a hydrocarbon compound present in the wastewater, for example. Microbial fuel cells (M...

Scott, Keith; Yu, Eileen Hao; Ghangrekar, Makarand Madhao; Erable, Benjamin; Duteanu, Narcis Mihai

2012-01-01

343

MOX fuel at BNFL  

International Nuclear Information System (INIS)

In 1989, BNFL decided to use the expertise developed for the Fast Reactor project to enter the thermal MOX fuels market with the aim of becoming a world leader in thermal MOX supply and to return the products from its reprocessing business to its customers as MOX fuel. To reach this objective the company developed a two-stage strategy which involved: (a) Constructing a small-scale plant, the MOX Demonstration Facility (MDF), on a short time-scale to produce commercial quality fuel for irradiation in commercial reactors, and (b) Constructing a small-scale plant, the Sellafield MOX Plant (SMP), for bulk fuel supply. MOX production in the MOX Demonstration Facility at Sellafield began in October 1993 and, since that time, the plant has produced more than 10 tonnes of MOX for BNFL's customers. The MDF was constructed to produce LWR MOX fuel, using BNFL's patented Short Binderless Route (SBR) in order to gain operational and irradiation experience to support fuel supply from the 120te/yr Sellafield MOX Plant (SMP). The first fuel from MDF was loaded into the Nordostschweizerische Kraftwerke (NOK) Beznau 1 reactor in July 1994 and since that time the plant has been used continuously to provide more fuel for NOK and other customers. Construction of the SMP commenced in April 1994 against a fast-track programme designed to have the plant producing its first MOX fuel by the end of 1997. The SMP will be the most flexible MOX fabrication plant in the world, capable of producing plant in the world, capable of producing PWR and BWR fuels using the SBR as the basis of the production process. (Author)

344

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

345

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

346

Alternate-Fueled Flight: Halophytes, Algae, Bio-, and Synthetic Fuels  

Science.gov (United States)

Synthetic and biomass fueling are now considered to be near-term aviation alternate fueling. The major impediment is a secure sustainable supply of these fuels at reasonable cost. However, biomass fueling raises major concerns related to uses of common food crops and grasses (some also called "weeds") for processing into aviation fuels. These issues are addressed, and then halophytes and algae are shown to be better suited as sources of aerospace fuels and transportation fueling in general. Some of the history related to alternate fuels use is provided as a guideline for current and planned alternate fuels testing (ground and flight) with emphasis on biofuel blends. It is also noted that lessons learned from terrestrial fueling are applicable to space missions. These materials represent an update (to 2009) and additions to the Workshop on Alternate Fueling Sustainable Supply and Halophyte Summit at Twinsburg, Ohio, October 17 to 18, 2007.

Hendricks, R. C.

2012-01-01

347

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

348

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

349

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

350

Nuclear fuel pellet loading apparatus  

International Nuclear Information System (INIS)

An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

351

Spent fuel workshop'2002  

International Nuclear Information System (INIS)

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO2 dissolution determined from electrochemical experiments with 238Pu doped UO2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with ? doped UO2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO2 / water interfaces under He2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of UO2(s): experimental approach and preliminary results on uranium oxide - water interface (J. Devoy), Preliminary results on studies on radiolysis effects on dissolution of UO2 (E. Ekeroth, M. Jonnson); Session 5 - Modeling of the Spent Fuel Dissolution: tUO2 dissolution and the effect of radiolysis (T. Lundstrom), Prediction of the effect of radiolysis (F. King), Experimental determination and chemical modeling of radiolytic processes at the spent fuel / water interface (E. Cera, J. Bruno, T. Eriksen, M. Grive, L. Duro); Session 6 - Influence of the Potential Evolution prior to the Water Access on IRF: Potential occurrence of ? self-irradiation enhanced-diffusion (H.J. Matzke, T. Petit), Are grain boundaries a stable microstructure? (Y. Guerin), Modeling RN instant release fractions from spent nuclear fuel under repository conditions (C.Poinssot, L. Johnson, P. Lovera). (J.S.)

352

MOX fuel assembly for PWR  

International Nuclear Information System (INIS)

In fuel assemblies for a PWR formed by bundling MOX fuel rods containing fissionable Pu, MOX fuel rods having a predetermined Pu enrichment degree are disposed at the outer circumferential portion except for each corner of the fuel assembly, and MOX fuel rods having a predetermined Pu enrichment degree which is higher than that of the fuel rods at the outer circumferential portion are disposed in the central portion of the fuel assembly. In addition, U fuel rods each containing gadolinium are disposed to each of the corners. This can reduce the kinds of the MOX fuel rods to two kinds without substantially reducing the loading amount of fissionable Pu as a whole of the assemblies while suppressing the power peeking of the fuel assemblies to the same level as that in conventional cases, thereby enabling to reduce fabrication cost upon manufacture of fuels. (T.M.)

353

Integral nuclear fuel element assembly  

International Nuclear Information System (INIS)

In a heavy water-moderated plutonium-uranium breeder reactor a low moderator-fuel ratio is desirable in order to increase the breeding ratio; however, fuel element spacing must be carefully controlled to provide good moderator flow and prevent the formation of hot spots on the fuel element surfaces. A fuel assembly made in accordance with this invention utilizes longitudinally-finned fuel pin cladding tubes arranged to form an integral fuel assembly by brazing together the continuous or interrupted fins of one fuel pin to the fins of other fuel pins. Alternatively, the fins of some fuel pins may be connected directly to the tubular section of other fuel pins. In another embodiment of this invention, the core is fabricated from a solid material having passages suited for coolant flow and fuel retention

354

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

355

Fuel assembly supporting structure  

International Nuclear Information System (INIS)

For use in forming the core of a pressurized-water reactor, a fuel assembly supporting structure for holding a bundle of interspaced fuel rods, is formed by interspaced end pieces having holes in which the end portions of control rod guide tubes are inserted, fuel rod spacer grids being positioned by these guide tubes between the end pieces. The end pieces are fastened to the end portions of the guide tubes, to integrate the supporting structure, and in the case of at least one of the end pieces, this is done by means which releases that end piece from the guide tubes when the end pieces receive an abnormal thrust force directed towards each other and which would otherwise place the guide tubes under a compressive stress that would cause them to buckle. The spacer grids normally hold the fuel rods interspaced by distances determined by nuclear physics, and buckling of the control rod guide tubes can distort the fuel rod spacer grids with consequent dearrangement of the fuel rod interspacing. A sudden loss of pressure in a pressurized-water reactor pressure vessel can result in the pressurized coolant in the vessel discharging from the vessel at such high velocity as to result in the abnormal thrust force on the end pieces of each fuel assembly, which could cause buckling of the control rod guide tubes when the end pieces are fixed to them in the normal rigid and unyielding manner

356

Inert matrix fuel (IMF)  

International Nuclear Information System (INIS)

The state of development and future perspectives of nuclear fuel with inert (non-fissionable) matrix for thermal and fast reactors are considered. Engineering requirements towards matrix materials for fuel pellets are formulated. The wet and dry processes of manufacturing the powder materials for matrices and fissionable compositions using co-deposition of powders following the nitrate technology and formation of microspheres by sol-gel method, as well as preparation of powders by mechanical treatment are described. It is shown that application of the fuel with inert matrix provides opportunities for sufficient increasing the utilization of available weapon plutonium, as well as the plutonium regenerated from spent uranium fuel, fuel burnup improvement, thorium involving into the fuel cycle and decreasing amounts of the fuel, which should be disposed. The conclusion is made that the oxide ceramics ZrO2, MgAl2O4, CeO2, Y3Al5O12, MgO are the most suitable for using as the materials for inert matrices

357

Nuclear fuel element  

International Nuclear Information System (INIS)

The present invention concerns a fuel element for a nuclear reactor, which prevents FP gases in a cladding tube from releasing into a cooling water upon rupture of fuels in a cladding tube by a back flow check valve disposed in the fuel element. Namely, in the fuel element of the present invention, fuels are contained in a cylindrical cladding tube having the upper and the lower ends sealed by end plugs respectively. A gas plenum vessel is disposed at the upper portion of the cladding tube. A back flow check valve is attached to the bottom of the gas plenum vessel. The back flow check valve is disposed in the direction of the FP gases flowing to the gas plenum in the gas plenum vessel. The back flow check valve has a ball valve system. With such a constitution, FP gas is transferred from the fuels to the gas plenum by the pressure difference during a normal state. However, when the cladding tube is ruptured, the back flow check valve is closed by the pressure difference. Accordingly, the FP gas in the gas plenum is not discharged out of the fuel elements. As the result, only the FP gas other than that in the gas plenum is discharged. (I.S.)

358

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.)

359

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.

360

Hydrogen vehicle fueling station  

Energy Technology Data Exchange (ETDEWEB)

Hydrogen fueling stations are an essential element in the practical application of hydrogen as a vehicle fuel, and a number of issues such as safety, efficiency, design, and operating procedures can only be accurately addressed by a practical demonstration. Regardless of whether the vehicle is powered by an internal combustion engine or fuel cell, or whether the vehicle has a liquid or gaseous fuel tank, the fueling station is a critical technology which is the link between the local storage facility and the vehicle. Because most merchant hydrogen delivered in the US today (and in the near future) is in liquid form due to the overall economics of production and delivery, we believe a practical refueling station should be designed to receive liquid. Systems studies confirm this assumption for stations fueling up to about 300 vehicles. Our fueling station, aimed at refueling fleet vehicles, will receive hydrogen as a liquid and dispense it as either liquid, high pressure gas, or low pressure gas. Thus, it can refuel any of the three types of tanks proposed for hydrogen-powered vehicles -- liquid, gaseous, or hydride. The paper discusses the fueling station design. Results of a numerical model of liquid hydrogen vehicle tank filling, with emphasis on no vent filling, are presented to illustrate the usefulness of the model as a design tool. Results of our vehicle performance model illustrate our thesis that it is too early to judge what the preferred method of on-board vehicle fuel storage will be in practice -- thus our decision to accommodate all three methods.

Daney, D.E.; Edeskuty, F.J.; Daugherty, M.A. [Los Alamos National Lab., NM (United States)] [and others

1995-09-01

 
 
 
 
361

Heating subsurface formations by oxidizing fuel on a fuel carrier  

Energy Technology Data Exchange (ETDEWEB)

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

362

Plasma sprayed and electrospark deposited zirconium metal diffusion barrier coatings  

International Nuclear Information System (INIS)

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.

363

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

364

Second International Conference on CANDU Fuel  

International Nuclear Information System (INIS)

Thirty-four papers were presented at this conference in sessions dealing with international experience and programs relating to CANDU fuel; fuel manufacture; fuel behaviour; fuel handling, storage and disposal; and advanced CANDU fuel cycles. (L.L.)

365

Fossil Fuels: Capstone  

Science.gov (United States)

This lesson summarizes our dependency upon fossil fuels, pointing out that there are very few aspects of our daily life that are not impacted by their use. The discussion centers around whether these fuels could be replaced and makes the point that there is a significant percentage of them which is used to manufacture products and is not simply burned for energy. The lesson includes an activity in which students use an online calculator to estimate how much of each fossil fuel they are responsible for consuming each year.

Pratte, John

366

Ionic liquids and fuels  

Energy Technology Data Exchange (ETDEWEB)

Ionic liquids have drawn a lot of attention in recent years due to their unique physical and chemical properties. They were successfully applied in various processes as catalysts, solvents, electrolytes, lubricants, thermo fluids or plasticizers. In this short review, we would like to show benefits ionic liquids can potentially bring to fuel problematic. The recent scientific development suggests that ionic liquids could be successfully applied in sulfur and mercury removal from hydrocarbons, serve as hypergolic fuels or play crucial role in the synthesis of alternative fuels. (orig.)

Adamova, Gabriela; Ahrens, Maria; Schubert, Thomas J.S. [IoLiTec GmbH, Heilbronn (Germany)

2013-06-01

367

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To provide nuclear fuel elements wherein gaps between the cladding tube and fuel pellets are filled with a metal powder, thereby to prevent occurrence of stress corrosion cracks in the cladding tube. Constitution: Nuclear fuel elements for use in LWR in which pellets of uranium dioxide are charged in the cladding tube made of zirconium, characterized in that gaps between said pellets of uranium dioxide and the cladding tube made of zircaloy are filled with a metal powder of at least a member selected from the group consisting of Zr, Cu, Ni, Fe, Sn, Mg, Ca and Zn. (Nakamura, S.)

368

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To reduce the amount of impurities equal to or less than that of crystal bar zirconium, thereby prevent stress corrosion cracks in a fuel can. Constitution: By applying electron beam welding to pure zirconium (sponge zirconium or forged pure zirconium), the total amount of impurities contained in zirconium such as Cu, Ca, Na, Mg, Ni, O and Fe is reduced to less than 0 - 1000 ppm, while the oxygen concentration is reduced to less than 200 - 400 ppm. The stress corrosion crack in the fuel can be prevented by using such zirconium to a zirconium liner layer in a composite type fuel can. (Takahashi, M.)

369

Reactor fuel rod  

International Nuclear Information System (INIS)

Purpose: To provided a fuel rod which incorporates a copper barrier tube improved in absorptibe power (getter effect) of fission product gases of iodine, cadmium or the like. Constitution: After a copper layer is formed on the inside surface of a zirconium alloy tube, the layer is once oxidezed, and an oxide layer thus produced is then reduced to form a fuel can. The copper layer thus formed exhibits very fine porous property resulting in remarkably large surface area. Therefore, the copper layer incorporates high absorptive power of fission product gases of iodine, cadmium or the like and restricts the excessive accumulation of these gases in a fuel rod. (Yoshihara, H.)

370

Nuclear reactor fuel unit  

International Nuclear Information System (INIS)

The object of this invention is to harness the force of the ascending coolant of the reactor to actuate a locking mechanism holding the fuel unit in position against the core support. The invention makes for easy unlocking of the unit so as to facilitate its removal. The fuel unit is held in position near its lower end by a compact device, of easy fabrication and use, that does not cause any significant drop in the pressure of the coolant in the reactor vessel. Under the invention, the fuel unit can be unlocked from a distance in the event of any failure of the normal locking components

371

Synthesis of fuel additives  

Energy Technology Data Exchange (ETDEWEB)

The great developments in engine design require the availability of high quality fuels with good low temperature properties. Pour point depressants and flow improvers have become of great assistance in fulfilling such demands. This paper is aimed at finding the optimum conditions to prepare some pour point depressants, flow improvers and ashless dispersant additives, which can be used in both gas oil and fuel oil samples. Accordingly, the work includes the following products: polysaccharides, allylnaphthalene succinimides, acylated diaromatics and esters of styrene-maleic anhydride copolyerms. Comparative evaluation of the synthesized products with available commercial additives showed their efficiency and suitability to use in fuels. (orig.)

Mohamed, M.M.; Abou el Naga, H.H.; El Meneir, M.F. (MISR Petroleum Co., Cairo (Egypt))

1999-01-01

372

Synthesis of fuel additives  

Energy Technology Data Exchange (ETDEWEB)

The great developments in engine design require the availability of high quality fuels with good low temperature properties. Pour point depressants and flow improvers have become of great assistance in fulfilling such demands. This paper is aimed at finding the optimum conditions to prepare some pour point depressants, flow improvers and ashless dispersant additives, which can be used in both gas oil and fuel oil samples. Accordingly, the work includes the following products: polysaccharides, allylnaphthalene succinimides, acylated diaromatics and esters of styrene-maleic anhydride copolyerms. Comparative evaluation of the synthesized products with available commercial additives showed their efficiency and suitability to use in fuels. (orig.)

Mohamed, M.M.; Abou el Naga, H.H.; El Meneir, M.F. [MISR Petroleum Co., Cairo (Egypt)

1999-11-01

373

Ammonia as a suitable fuel for fuel cells  

Directory of Open Access Journals (Sweden)

Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

ShanwenTao

2014-08-01

374

Ammonia as a suitable fuel for fuel cells  

Digital Repository Infrastructure Vision for European Research (DRIVER)

Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel c...

ShanwenTao

2014-01-01

375

Fuels and combustion technologies for power generation. Nuclear fuel cycle  

Energy Technology Data Exchange (ETDEWEB)

As nuclear fuel cycle is to use nuclear fuel materials through in circulation, by which it is a striking characteristics impossible to find in fossil fuels such as petroleum, coal, and so on to conduct reuse and effective use of fuels. In the fuel cycle, a path to use nuclear fuel materials in a reactor is called its up-stream, and another path after taking off from the reactor is called its down stream. Here was outlined on these nuclear fuel cycles under containing its trends in and out of Japan. (G.K.)

Fujita, Chitoshi [Japan Atomic Power Co., Tokyo (Japan); Nakajima, Isao; Taga, Junichi; Izutsu, Sadayuki

1999-10-01

376

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

377

Features of fuel performance at high fuel burnups  

International Nuclear Information System (INIS)

Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

378

Seventh Edition Fuel Cell Handbook  

Energy Technology Data Exchange (ETDEWEB)

Provides an overview of fuel cell technology and research projects. Discusses the basic workings of fuel cells and their system components, main fuel cell types, their characteristics, and their development status, as well as a discussion of potential fuel cell applications.

NETL

2004-11-01

379

Automotive Fuel and Exhaust Systems.  

Science.gov (United States)

Materials are provided for a 14-hour course designed to introduce the automotive mechanic to the basic operations of automotive fuel and exhaust systems incorporated on military vehicles. The four study units cover characteristics of fuels, gasoline fuel system, diesel fuel systems, and exhaust system. Each study unit begins with a general…

Irby, James F.; And Others

380

Fuel Injector With Shear Atomizer  

Science.gov (United States)

Atomizer for injecting liquid fuel into combustion chamber uses impact and swirl to break incoming stream of fuel into small, more combustible droplets. Slanted holes direct flow of liquid fuel to stepped cylindrical wall. Impact on wall atomizes liquid. Air flowing past vanes entrains droplets of liquid in swirling flow. Fuel injected at pressure lower than customarily needed.

Beal, George W.; Mills, Virgil L.; Smith, Durward B., II; Beacom, William F.

1995-01-01

 
 
 
 
381

Mechanical decladding of fuel elements  

International Nuclear Information System (INIS)

Mechanical decladding and desintegration of irradiated fuel elements is not a problem any more. There are now methods at hand enabling the extensive preliminary desintegration of fuel elements, i.e. singulizing of fuel rods, by the use of tested rod row cutting machine or by cutting across the entire fuel element cross section. (orig./DG)

382

Spent fuel storage and isolation  

International Nuclear Information System (INIS)

The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

383

Advanced Fuels Campaign 2012 Accomplishments  

Energy Technology Data Exchange (ETDEWEB)

The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

Not Listed

2012-11-01

384

Bioethanol: Fueling sustainable transportation  

Energy Technology Data Exchange (ETDEWEB)

Ethanol made from biomass, or bioethanol, can positively impact the national energy security, the economy, and the environment. Producing and using bioethanol can help alleviate some of the negative impacts of the dependence on fossil fuels.

Neufeld, S.

2000-05-25

385

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

386

The spent fuel fate  

International Nuclear Information System (INIS)

The spent fuel is not a waste. It can be upgrade by a reprocessing which extracts all products able to produce energy. The today situation is presented and economically analyzed and future alternatives are discussed. (A.L.B.)

387

Fuel cycle safeguards approaches  

International Nuclear Information System (INIS)

A review is presented of the status of studies recently performed in the United States of America on alternative safeguards approaches for light water reactor fuel cycles under the safeguards system of the Non-Proliferation Treaty; and a new method is described for consistent allocation of limited inspection manpower resources. The studies concern the advantages and disadvantages of using safeguards approaches for groups of nuclear facilities that interact to form a nuclear fuel cycle, rather than an approach that is facility specific, treating each facility as an isolated entity. The allocation methodology makes use of inspection evaluation criteria used by the IAEA, as well as a procedure to provide consistency in allocation of limited resources to individual facilities of any fuel cycle while considering all other facilities in the fuel cycle. (author)

388

Fuel gas from biodigestion  

Science.gov (United States)

Biodigestion apparatus produces fuel gas (primarily methane) for domestic consumption, by anaerobic bacterial digestion of organic matter such as aquatic vegetation. System includes 3,786-1 cylindrical container, mechanical agitator, and simple safe gas collector for short term storage.

Mcdonald, R. C.; Wolverton, B. C.

1979-01-01

389

Spent fuel storage racks  

International Nuclear Information System (INIS)

Purpose: To decrease a spent fuel storage area occupied in a storage pool by increasing the storage density of the spent fuel, thereby enabling semi-permanent storage. Constitution: A framework of a spent fuel storage racks is made of stainless steel; and cells for holding fuel rod assemblies are formed by fitting, in its wall surface, ceramic tiles produced by sintering a mixture of such medium materials as Al2O3, SiO2, and ZrO2 and such elements as Eu, Hf, Gd, Sm, Co and Rh used as neutron absorbers which emit gamma rays in the process of reaction to neutrons. The most desirable neutron absorber to be used is Hf which has a high neutron absorbing capacity and a long half-life, is capable of keeping on absorbing neutrons even after neutron absorption, and makes (n,?) neutron absorption reaction, emitting gamma rays without accompanying any change of the ceramic tiles with time. (Sekiya, K.)

390

Spent fuel storage rack  

International Nuclear Information System (INIS)

Purpose: To increase fuel storage density by directly connecting angle cylinders of boron-added austenite steel materials with each other in both longitudinal and lateral directions. Constitution: The cells of the angle cylinders only are constructed by directly connecting with each other the angle cylinders of highly accurate dimensions composed of boron-added (1 % or less) austenite stainless-steel materials on the base which has the coolant flow paths fixed to the bottom of the fuel storage pool. With this, the inside dimensions of each cell can be constructed with high accuracy, and at the same time, the angle cylinder rigidity can be prevented from being lowered. Each cell is designed with a little larger dimensions than those of the fuel. As the inside dimensions and the pitches of each cell can be minimized, the fuel storage density can be increased as much. (Takahashi, M.)

391

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.

392

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

393

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

394

Integral-fuel blocks  

International Nuclear Information System (INIS)

A prismatic moderator block is described which has fuel-containing channels and coolant channels disposed parallel to each other and to edge faces of the block. The coolant channels are arranged in rows on an equilateral triangular lattice pattern and the fuel-containing channels are disposed in a regular lattice pattern with one fuel-containing channel between and equidistant from each of the coolant channels in each group of three mutually adjacent coolant channels. The edge faces of the block are parallel to the rows of coolant channels and the channels nearest to each edge face are disposed in two rows parallel thereto, with one of the rows containing only coolant channels and the other row containing only fuel-containing channels. (Official Gazette)

395

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

396

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

397

Nuclear fuel pellet  

International Nuclear Information System (INIS)

In a nuclear fuel pellet comprising uranium dioxide or an oxide mixture of uranium dioxide and plutonium dioxide as a main ingredient, additives which are not easily solid-solubilized into nuclear fuels are dispersed uniformly in the nuclear fuel pellet. With such a constitution, they provide thermodynamically stable places like that crystal boundary, i.e., sinks, for fission products not solid-soluble to the matrix of the nuclear fuel material. Therefore, most of the fission products formed in the crystal boundary and not solid-soluble to the matrix moves to the nearer sink at the boundary between fine particles of the additives and the matrix. Since fission products of rare gases are accumulated on the boundary and the amount of the fission products moved to the crystal boundary is negligible, most of them is kept in the pellet. With such a constitution, stress corrosion cracks and liquid metal embrittlement are scarcely caused to the cladding tube. (T.M.)

398

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To shield gamma rays irradiated through nuclear reaction of burnable poisons such as Gd2O3 thereby suppress the increase in the oxygen concentration due to radiolysis of reactor water in direct contact with the outer surface of zircalloy cladding tubes. Constitution: In a nuclear fuel element containing burnable poisons for use in light water-cooled nuclear reactor, heat conductive gamma ray shielding materials are put between nuclear fuel pellets and a zircalloy cladding tubes. Further, the fuel element is structured such that molten UO2 particles are dispersed like that of cermet in sintered pellets. Gamma rays generated form Gd are shielded by the molten UO2, whereby the radiolysis of reactor water near the surface of the zircalloy cladding tube is reduced as compared with that in the ordinary fuel element, to suppress the increase in the oxygen concentration. As the result, the corrosion at the outer surface of the zircalloy cladding tube can be decreased. (Nakamoto, H.)

399

Automatic fuel exchanging device  

International Nuclear Information System (INIS)

Purpose: To enable to designate the identification number of a fuel assembly in a nuclear reactor pressure vessel thereby surely exchanging the designated assembly within a short time. Constitution: Identification number (or letter) pressed on a grip of a fuel assembly is to be detected by a two-dimensional ultrasonic probe of a pull-up mechanism. When the detected number corresponds with the designated number, a control signal is outputted, whereby the pull-up drive control mechanism or pull-up mechanism responds to pull-up and exchange the fuel assembly of the identified number. With such a constitution, the fuel assembly can rapidly and surely be recognized even if pressed letters deviate to the left or right of the probe, and further, the hinge portion and the signal processing portion can be simplified. (Horiuchi, T.)

400

PEM regenerative fuel cells  

Science.gov (United States)

This paper will update the progress in developing electrocatalyst systems and electrode structures primarily for the positive electrode of single-unit solid polymer proton exchange membrane (PEM) regenerative fuel cells. The work was done with DuPont Nafion 117 in complete fuel cells (40 sq cm electrodes). The cells were operated alternately in fuel cell mode and electrolysis mode at 80 C. In fuel cell mode, humidified hydrogen and oxygen were supplied at 207 kPa (30 psi); in electrolysis mode, water was pumped over the positive electrode and the gases were evolved at ambient pressure. Cycling data will be presented for Pt-Ir catalysts and limited bifunctional data will be presented for Pt, Ir, Ru, Rh, and Na(x)Pt3O4 catalysts as well as for electrode structure variations.

Swette, Larry L.; Laconti, Anthony B.; Mccatty, Stephen A.

1993-01-01

 
 
 
 
401

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

402

Internal fuel pin oxidizer  

International Nuclear Information System (INIS)

A nuclear fuel pin has positioned within it material which will decompose to release an oxidizing agent which will react with the cladding of the pin and form a protective oxide film on the internal surface of the cladding

403

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

404

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

405

Improved CANDU fuel performance  

International Nuclear Information System (INIS)

The fuel defect rate in CANDU power reactors has been very low (0.06 percent) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness and wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

406

Safety of fuel using  

International Nuclear Information System (INIS)

The research reactor complex IR-100 comprises the 200 KW power nuclear reactor and critical assembly. The reactor fuel rods operate since 1967. The tightness of fuel rod casings now has been determined by using of the coolant radiochemical analysis. Prognosis of the casings tightness change has been carried out as well. The minimum and maximum periods of the fuel rods operation in the future has been determined by the graphic method. In the critical assembly fuel rods the long-lived radioactive fission products are accumulated with time. There is the risk of irradiation of IR-100 personnel and students. The prognosis of the long-lived radioactive isotopes accumulating and increase of emanation has been carried out by graphic-analytical method. (author)

407

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

408

Spent fuel reprocessing options  

International Nuclear Information System (INIS)

The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the development of advanced nuclear fuel cycles is also given. As a conclusion, spent fuel reprocessing options have evolved significantly since the start of nuclear energy application. There is a large body of industrial experience in fuel cycle technologies complemented by research and development programs in several countries

409

Increasing TRIGA fuel lifetime with 12 wt.% U TRIGA fuel  

International Nuclear Information System (INIS)

In-core fuel management studies have been performed for the Penn State Breazeale Reactor (PSBR) wherein 12 wt % U fuel elements are used to replace the standard 8.5 wt % U TRIGA fuel. The core configuration used to develop a calculational model was a 90-element hexagonal array, which is representative of the PSBR core, and consists of five hexagonal rings surrounding a central thimble containing water. The technique employed for refueling the core fully loaded with 8.5 wt % U fuel involves replacing 8.5 wt % U fuel with 12 wt % U fuel using an in-out reloading scheme. A batch reload consists of 6 new 12 wt % U fuel elements. Placing the 12 wt % U fuel in the B ring produces fuel temperatures (?450 oC) that are well below the 800oC maximum limitation when the PSBR is operating at its maximum allowed power of 1 Megawatt. The advantages of using new 12 wt % U fuel to replace the burned up 8.5 wt % U fuel in the B ring over refueling strictly with 8.5 wt % U-Zr TRIGA fuel are clearly delineated in Table 1 where cost calculations used the General Atomic pre-1972 prices for TRIGA fuel, i.e., $1500 and $1650 for an 8.5 and 12 wt % U fuel element, respectively. Experimental results obtained to date utilizing the 12 wt % U fuel elements agree with the computed results. (author)

410

Fuel rod behaviour at high burnup WWER fuel cycles  

International Nuclear Information System (INIS)

The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

411

Direct Methanol Fuel Cell, DMFC  

Digital Repository Infrastructure Vision for European Research (DRIVER)

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 therefor...

Amornpitoksuk, P.

2003-01-01

412

WWER-1000 fuel cycle improvement  

International Nuclear Information System (INIS)

The problems of organization of fuel cycles with different operation time of stationary load for the reactor WWER-1000 are considered. The outcomes of matching of the characteristics for stationary load constructed on fuel cells of existing and improved designs are presented. Improved designs of a fuel cell are include increase of an altitude of a fuel stake, change of outside and axial diameters of a fuel pellet, change thickness of a cladding of a fuel cell. Effect of the layout solutions on improving of a fuel cycle WWER-1000 also considered (Authors)

413

Fuels for Transportation  

Digital Repository Infrastructure Vision for European Research (DRIVER)

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

414

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)

415

A fuel pyrolysis method  

Energy Technology Data Exchange (ETDEWEB)

In the method for pyrolysis (Pz) of fuels (Tp) through passing them through a melted heat carrier (TN), in order to increase the process effectiveness, the fuels to be pyrolyzed are fed into the melted heat carrier in the form of a sharp stream. The balancing of the temperature of the heat carrier in the heating chambers and the pyrolysis chambers occurs due to mixing of its weight by the sharp stream and through convection.

Chukhanov, Z.F.; Dvoskin, G.I.; Kashurichev, A.P.; Khmelevskaya, Ye.D.; Kurochkin, A.I.; Pulkina, M.K.

1982-01-01

416

Fuel gas conditioning process  

Science.gov (United States)

A process for conditioning natural gas containing C.sub.3+ hydrocarbons and/or acid gas, so that it can be used as combustion fuel to run gas-powered equipment, including compressors, in the gas field or the gas processing plant. Compared with prior art processes, the invention creates lesser quantities of low-pressure gas per unit volume of fuel gas produced. Optionally, the process can also produce an NGL product.

Lokhandwala, Kaaeid A. (Union City, CA)

2000-01-01

417

Biodegradation of biodiesel fuels  

International Nuclear Information System (INIS)

Biodiesel fuel test substances Rape Ethyl Ester (REE), Rape Methyl Ester (RME), Neat Rape Oil (NR), Say Methyl Ester (SME), Soy Ethyl Ester (SEE), Neat Soy Oil (NS), and proportionate combinations of RME/diesel and REE/diesel were studied to test the biodegradability of the test substances in an aerobic aquatic environment using the EPA 560/6-82-003 Shake Flask Test Method. A concurrent analysis of Phillips D-2 Reference Diesel was also performed for comparison with a conventional fuel. The highest rates of percent CO2 evolution were seen in the esterified fuels, although no significant difference was noted between them. Ranges of percent CO2 evolution for esterified fuels were from 77% to 91%. The neat rape and neat soy oils exhibited 70% to 78% CO2 evolution. These rates were all significantly higher than those of the Phillips D-2 reference fuel which evolved from 7% to 26% of the organic carbon to CO2. The test substances were examined for BOD5 and COD values as a relative measure of biodegradability. Water Accommodated Fraction (WAF) was experimentally derived and BOD5 and COD analyses were carried out with a diluted concentration at or below the WAF. The results of analysis at WAF were then converted to pure substance values. The pure substance BOD5 and COD values for test substances were then compared to a control substance, Phillips D-2 Reference fuel. No significant difference was notce fuel. No significant difference was noted for COD values between test substances and the control fuel. (p > 0.20). The D-2 control substance was significantly lower than all test substances for BCD, values at p 5 value

418

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

419

IFR fuel cycle  

Energy Technology Data Exchange (ETDEWEB)

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.

Battles, J.E.; Miller, W.E. (Argonne National Lab., IL (United States)); Lineberry, M.J.; Phipps, R.D. (Argonne National Lab., Idaho Falls, ID (United States))

1992-01-01

420

IFR fuel cycle  

Energy Technology Data Exchange (ETDEWEB)

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.

Battles, J.E.; Miller, W.E. [Argonne National Lab., IL (United States); Lineberry, M.J.; Phipps, R.D. [Argonne National Lab., Idaho Falls, ID (United States)

1992-04-01

 
 
 
 
421

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

422

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

423

SRC Residual fuel oils  

Science.gov (United States)

Coal solids (SRC) and distillate oils are combined to afford single-phase blends of residual oils which have utility as fuel oils substitutes. The components are combined on the basis of their respective polarities, that is, on the basis of their heteroatom content, to assure complete solubilization of SRC. The resulting composition is a fuel oil blend which retains its stability and homogeneity over the long term.

Tewari, Krishna C. (Whitehall, PA); Foster, Edward P. (Macungie, PA)

1985-01-01

424

Fuel cell system configurations  

International Nuclear Information System (INIS)

Fuel cell stack configurations are disclosed that have elongated polygonal crosssectional shapes and gaskets at the peripheral faces to which flow manifolds are sealingly affixed. Process channels convey a fuel and an oxidant through longer channels, and a cooling fluid is conveyed through relatively shorter cooling passages. The polygonal structure preferably includes at least two right angles , and the faces of the stack are arranged in opposite parallel pairs

425

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

2012-09-14

426

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

427

Diesel Dual Fuel Technology  

Digital Repository Infrastructure Vision for European Research (DRIVER)

The work covered in this report includes a literature study of the diesel dual fuel technology (hereinafter referred to as DDF). The literature study covers earlier work that has been done regarding the use of compressed natural gas (CNG) and diesel in a DDF system. Moreover a one cylinder research diesel engine was fitted with a CNG fuel system for experimental testing. The installation of the engine and the test cell is described. Results from the tests are presented covering emissions, hea...

Renman, Emil

2008-01-01

428

Fuel injection system  

Energy Technology Data Exchange (ETDEWEB)

A fuel injection system for an internal combustion engine includes electromagnetic injection valves controlled by a fuel control unit which receives signals from a camshaft actuated switch, a position-dependent throttle transducer and an oxygen sensor. When the oxygen sensor changes output levels, the transmission of this information is delayed, by the action of a switching transistor controlled by a monostable multivibrator, for a period of time equal to the internal time constant of the multivibrator.

Herth, H.; Kraus, B.; Sautter, W.; Wessel, W.

1983-03-15

429

Gas as marine fuel  

Digital Repository Infrastructure Vision for European Research (DRIVER)

This Bachelor’s thesis examined the technical possibilities of using natural gas as marine fuel for merchant fleet. The main aim was to gather all relevant information on how natural gas can be used in maritime sector. Due to the protection of environment, remarkable numbers of re-quirements are set for maritime transport sector regarding the combustion of fossil fuels. The aim is to reduce or even eliminate harmful emissions. Among other options liquefied natural gas may be environment...

Vabar, Tarvi

2011-01-01

430

EPRI fuel cladding integrity program  

Energy Technology Data Exchange (ETDEWEB)

The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

1997-01-01

431

EPRI fuel cladding integrity program  

International Nuclear Information System (INIS)

The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R ampersand D includes fuel and cladding. In this paper, only R ampersand D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response

432

Decladding method of spent fuel  

International Nuclear Information System (INIS)

The present invention is used for a decladding step in a reprocessing step for spent nuclear fuels. Namely, notches are formed to a spent nuclear fuel rod containing fuel pellets in a cladding tube. Then, if load is applied thereto, the fuel rods are cut easily and reliably at the notched portions. The cut fuel rods are heated at an air atmosphere at from 500degC to 600degC. Then, fuel pellets in the fuel rod are oxidized to cause expansion of the volume. As a result, fuel pellets and the cladding tube can be separated reliably. Heretofore, spent fuels have been finely sheared, and the fuel pellets have been melted and extracted, but this method involves a drawback that the shearing blades are vigorously exhausted, and the waste shearing blades form radioactive wastes. In the present invention, the exhaustion of tools and wastes to be generated can be reduced. (I.S.)

433

Thermal breeder fuel enrichment zoning  

International Nuclear Information System (INIS)

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect. 1 figure

434

Overview of fuel conversion  

International Nuclear Information System (INIS)

The conversion of solid fuels to cleaner-burning and more user-friendly solid liquid or gaseous fuels spans many technologies. In this paper, the authors consider coal, residual oil, oil shale, tar sends tires, municipal oil waste and biomass as feedstocks and examine the processes which can be used in the production of synthetic fuels for the transportation sector. The products of mechanical processing to potentially usable fuels include coal slurries, micronized coal, solvent refined coal, vegetable oil and powdered biomall. The thermochemical and biochemical processes considered include high temperature carbide production, liquefaction, gasification, pyrolysis, hydrolysis-fermentation and anaerobic digestion. The products include syngas, synthetic natural gas, methanol, ethanol and other hydrocarbon oxygenates synthetic gasoline and diesel and jet engine oils. The authors discuss technical and economic aspects of synthetic fuel production giving particular attention and literature references to technologies not discussed in the five chapters which follow. Finally the authors discuss economic energy, and environmental aspects of synthetic fuels and their relationship to the price of imported oil

435

Nalco Fuel Tech  

Energy Technology Data Exchange (ETDEWEB)

The Nalco Fuel Tech with its seat at Naperville (near Chicago), Illinois, is an engineering company working in the field of technology and equipment for environmental protection. A major portion of NALCO products constitute chemical materials and additives used in environmental protection technologies (waste-water treatment plants, water treatment, fuel modifiers, etc.). Basing in part on the experience, laboratories and RD potential of the mother company, the Nalco Fuel Tech Company developed and implemented in the power industry a series of technologies aimed at the reduction of environment-polluting products of fuel combustion. The engineering solution of Nalco Fuel Tech belong to a new generation of environmental protection techniques developed in the USA. They consist in actions focused on the sources of pollutants, i.e., in upgrading the combustion chambers of power engineering plants, e.g., boilers or communal and/or industrial waste combustion units. The Nalco Fuel Tech development and research group cooperates with leading US investigation and research institutes.

Michalak, S.

1995-12-31

436

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)

437

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

438

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.)

439

Nuclear fuel elements  

International Nuclear Information System (INIS)

Purpose: To prevent stress corrosion cracks by retaining iodine as much as possible in fuel pellets and stabilizing iodine released out of the pellets so that it may not give chemical effects on zirconium alloy fuel cans. Constitution: UO2 fuel powder is incorporated and mixed with cesium by an amount in excess of solid solution limit, molded under pressure and then sintered, whereby the cesium is crystallized out in the crystal grain boundary to form polycrystalline body containing numerous crystal grains therein. When such fuel pellets are assembled into fuel elements and used in reactors, iodine yielded through fission in the grains diffuses and reacts with cesium to produce stable CsI. If CsI is released from the grain boundary out of the pellets, it does not react with zirconium alloy fuel cans since its free formation energy is lower than the free formation energy for zirconium iodides. Cesium may be replaced with K, Rb, Na, Li, Ca and the like. (Horiuchi, T.)

440

Fuel handling benchmarking  

International Nuclear Information System (INIS)

On-power fuelling is unique to the CANDU type of reactor. The systems and equipment used to handle the fuel from the time it enters the station to the time it is transferred to the spent fuel bay are designed, operated and maintained exclusively for the CANDU stations. Over the last ten years it was perceived by several CANDU utility executives and outside organizations that CANDU fuel handling (FH) performance was degrading. FH organizations were seen as insular from the rest of the station and did not appear to be working to the same standards of excellence as the rest of the industry. The concerns raised were common to the industry. In 2005, COG was requested by one of its members to undertake an industry wide fuel handling Benchmarking (FHB) exercise of CANDU fuel handling organizations. The COG members decided to 'Take the cape off fuel handling' allowing all CANDU stations to see: actual performance of FH organizations; i.e. based on performance not perception, FH best practices, and identification of stations with best practices available for widespread use. All COG members joined COG project JP 4207. Taken together, the FH Benchmarking Final Report and the station Reports provide a good picture of current CANDU FH best practices and performance. (author)