WorldWideScience

Sample records for average fuel enrichment

  1. Guide for the estimation of the α and β coefficients in the Average enrichment equation as burnt function by fuel type

    International Nuclear Information System (INIS)

    The objective of the report is to determine manually or by means of a calculation sheet, the coefficients α and β of the average enrichment equation as function of the fuel burnt (B) using the Lineal Reactivity Pattern, with information generated by the RECORD code of the FMS package. (Author)

  2. 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

  3. 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

  4. Hydrogen-enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  5. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    In the 1950s and 1960s, low-power research reactors were built around the world utilized MTR-type fuel elements containing 20% enriched uranium. However, the demand for higher specific power created a need for greater uranium-235 concentrations. Early difficulties in increasing uranium content led to the substitution of highly enriched uranium in place of the 20% enriched fuel previously utilized. The highly enriched material also yielded other benefits including longer core residence time, higher specific reactivity, and somewhat lower cost. Highly enriched material then became readily available and was used for high-power reactors as well as in low-power reactors where 20% enriched material would have sufficed. The trend toward higher and higher specific power also led to the development of the dispersion-type fuels which utilized highly enriched uranium at a concentration of about 40 wt%. In the 1970's, however, concerns were raised about the proliferation resistance of fuels and fuel cycles. As a consequence, the U.S. Department of State has recently prohibited the foreign shipment of highly enriched material, except where prior contractual obligation or special merit exists. This will impact on the availability and utilization of highly enriched uranium for research and test reactor fuel. It has also stimulated development programs on fuels with higher uranium content which would allow the use of uranium of lower enrichment. The purpose of this report is to briefly describe the overall fuel development program which is coordinated by Argonne National Laboratory for the Department of Energy, and to indicate the current and potential uranium loadings. Other reports will address the individual fuel-development activities in greater detail

  6. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    In the 1950s and 1960s, low-power research reactors were built around the world utilized MTR-type fuel elements containing 20% enriched uranium. However, the demand for higher specific power created a need for greater uranium-235 concentrations. Early difficulties in increasing uranium content led to the substitution of highly enriched uranium in place of the 20% enriched fuel previously utilized. The highly enriched material also yielded other benefits including longer core residence time, higher specific reactivity, and somewhat lower cost. Highly enriched material then became readily available and was used for high-power reactors as well as in low-power reactors where 20% enriched material would have sufficed. The trend toward higher and higher specific power also led to the development of the dispersion-type fuels which utilized highly enriched uranium at a concentration of about 40 wt%. In the 1970's, however, concerns were raised about the proliferation resistance of fuels and fuel cycles. As a consequence, the U.S. Department of State has recently prohibited the foreign shipment of highly enriched material, except where prior contractual obligation or special merit exists. This will impact on the availability and utilization of highly enriched uranium for research and test reactor fuel. It has also stimulated development programs on fuels with higher uranium content which would allow the use of uranium of lower enrichment. The purpose of this report is to briefly describe the overall fuel-development program which is coordinated by Argonne National Laboratory for the Department of Energy, and to indicate the current and potential uranium loadings. Other reports will address the individual fuel-development activities in greater detail

  7. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  8. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  9. Optimal axial enrichment distribution of the boiling water reactor fuel under the Haling strategy

    International Nuclear Information System (INIS)

    The axial enrichment distribution of boiling water reactor fuel is optimized to improve uranium utilization subject to constraints on thermal margins. It is assumed that the reactor is operated under the Haling strategy, so that determination of the enrichment distribution can be decoupled from the poison management. This nonlinear optimization problem is solved using a method of approximation programming, where each iteration step is formulated in terms of linear goal programming to handle infeasible problems. The core is represented by an axial one-dimensional model. The average enrichment of a two-region fuel can be slightly reduced by increasing the enrichment of the lower half rather than the upper half. The optimal solutions for a 24-region fuel, in which the enrichments of indivdual nodes can differ from one another, display double-humped enrichment distributions. The natural uranium blanket design is also investigated, and it is concluded that blanketed fuel is practically optimal using the Haling strategy

  10. Cycle Average Peak Fuel Temperature Prediction Using CAPP/GAMMA+

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam-il; Lee, Hyun Chul; Lim, Hong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In order to obtain a cycle average maximum fuel temperature without rigorous efforts, a neutronics/thermo-fluid coupled calculation is needed with depletion capability. Recently, a CAPP/GAMMA+ coupled code system has been developed and the initial core of PMR200 was analyzed using the CAPP/GAMMA+ code system. The GAMMA+ code is a system thermo-fluid analysis code and the CAPP code is a neutronics code. The General Atomics proposed that the design limit of the fuel temperature under normal operating conditions should be a cycle-averaged maximum value. Nonetheless, the existing works of Korea Atomic Energy Research Institute (KAERI) only calculated the maximum fuel temperature at a fixed time point, e.g., the beginning of cycle (BOC) just because the calculation capability was not ready for a cycle average value. In this work, a cycle average maximum fuel temperature has been calculated using CAPP/GAMMA+ code system for the equilibrium core of PMR200. The CAPP/GAMMA+ coupled calculation was carried out for the equilibrium core of PMR 200 from BOC to EOC to obtain a cycle average peak fuel temperature. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle average peak fuel temperature was calculated as 1181 .deg. C, which is below the design target of 1250 .deg. C.

  11. Development of long-life low enrichment fuel

    International Nuclear Information System (INIS)

    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 non-proliferation 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 this year, 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

  12. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Science.gov (United States)

    2010-10-01

    ... subpart F of 40 CFR part 600. (c) Average fuel economy. Average fuel economy must be based upon fuel... under paragraph (c) of this section and has been determined and approved under 40 CFR part 600, the...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not...

  13. RERTR program progress in qualifying reduced-enrichment fuels

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J.L.

    1982-01-01

    In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the US Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of minature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given.

  14. Fuel optimum low-thrust elliptic transfer using numerical averaging

    Science.gov (United States)

    Tarzi, Zahi; Speyer, Jason; Wirz, Richard

    2013-05-01

    Low-thrust electric propulsion is increasingly being used for spacecraft missions primarily due to its high propellant efficiency. As a result, a simple and fast method for low-thrust trajectory optimization is of great value for preliminary mission planning. However, few low-thrust trajectory tools are appropriate for preliminary mission design studies. The method presented in this paper provides quick and accurate solutions for a wide range of transfers by using numerical orbital averaging to improve solution convergence and include orbital perturbations. Thus, preliminary trajectories can be obtained for transfers which involve many revolutions about the primary body. This method considers minimum fuel transfers using first-order averaging to obtain the fuel optimum rates of change of the equinoctial orbital elements in terms of each other and the Lagrange multipliers. Constraints on thrust and power, as well as minimum periapsis, are implemented and the equations are averaged numerically using a Gausian quadrature. The use of numerical averaging allows for more complex orbital perturbations to be added in the future without great difficulty. The effects of zonal gravity harmonics, solar radiation pressure, and thrust limitations due to shadowing are included in this study. The solution to a transfer which minimizes the square of the thrust magnitude is used as a preliminary guess for the minimum fuel problem, thus allowing for faster convergence to a wider range of problems. Results from this model are shown to provide a reduction in propellant mass required over previous minimum fuel solutions.

  15. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

    In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. This fuel development effort included work to increase the density of fuels which were being used at the time the Program began and work on a fuel with the potential for much higher density. The ultimate goal of the fuel development and testing phase of the Program is to 'qualify' the fuel for use. A fuel is considered qualified when a sufficient data base for the fuel exists that it can be approved by regulating bodies for use in reactors. To convert a core to the use of reduced-enrichment fuel it is necessary to show that the core will behave properly during normal and off-normal operating conditions and to show that the fuel will behave properly to a reasonable margin beyond the conditions expected during normal operation. It is this latter area that this paper will address. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of miniature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data reported in other papers presented during this meeting and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given

  16. Delays hit conversion to low enriched uranium fuel

    Science.gov (United States)

    Allen, Michael

    2016-03-01

    Eliminating highly enriched uranium (HEU) fuel from civilian research reactors around the world will take a lot longer than anticipated, according to a new study by the US National Academies of Sciences, Engineering and Medicine.

  17. Development of low enrichment MTR fuel at Dounreay

    International Nuclear Information System (INIS)

    Work up to October 1983 on the development of a manufacturing route for the manufacture of low enriched fuel at Dounreay concentrated on the roll-bonding method of plate manufacture. Both U-Al alloy and U3O8-Al cermet elements at 45% enrichment have been irradiated and the fabrication of 20% enriched U3O8-Al cermet elements is in hand. (author). 3 refs, 2 tabs

  18. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  19. Feasibility of Low Enriched Uranium Fuel for Space Nuclear Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    The purpose of this initial study is to create a baseline with which to perform further analysis and to build a solid understanding of the neutronic characteristics of a solid core for the nuclear thermal rocket. Once consistency with work done at Idaho National Laboratory (INL) is established, this paper will provide a study of other fuel types, such as low and medium-enriched uranium fuels. This paper will examine how the implementation of each fuel type affects the multiplication factor of the reactor, and will then explore different possibilities for alterations needed to accommodate their successful usage. The reactor core analysis was done using the MCNP5 code. While this study has not shown that the SNRE can be easily retrofitted for low-enriched U fuel, it has made a detailed study of the SNRE, and identified the difficulties of the implementation of low-enriched fuels in small nuclear rockets. These difficulties are the need for additional moderation and fuel mass in order to achieve a critical mass. Neither of these is insurmountable. Future work includes finding the best method by which to increase the internal moderation of the reactor balanced with appropriate sizing to prevent neutron leakage. Both of these are currently being studied. This paper will present a study of the Small Nuclear Rocket Engine (SNRE) and the feasibility of using low enriched Uranium (LEU) instead of the traditional high enriched Uranium (HEU) fuels.

  20. Richland five-year O2 R and D Program: Enriched fuel processing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-06-30

    In anticipation of a continuing trend for reductions in military plutonium requirements, reactor programs at the Richland site are being directed toward other long-term production objectives. For the general case of an alternate reactor product, uranium-238 would be displaced from the reactor by another target material, e.g., lithium (for tritium production) or neptunium-237 (for plutonium-238 production), and the remaining fuel would require higher enrichment (increase of uranium-235 concentration) to maintain reactor reactivity. The operating fuel reprocessing facilities at Hanford were originally designed for the processing of fuels containing less than one percent U-235 (pre-irradiation basis). Today, limited amounts of a ``spike`` fuel, averaging about 1.15 percent U-235, are included in the production load, and demonstration quantities of 2.1 percent enriched coproduct fuels have been processed under special test support conditions. Anticipated reactor programs requiring higher enrichment fuels pose new problems of reprocessing technology. These problems have their bases in the increased U-235 content of the fuel, and in the material and design features provided to obtain a higher specific power in the reactor. The programs required to develop the technological bases for reprocessing proposed Hanford fuels of greater enrichments, generally in excess of one percent U-235, are described by this document

  1. 40 CFR 600.510-86 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy... Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy § 600.510-86 Calculation of average fuel economy. (a) Average fuel economy will be calculated to the...

  2. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  3. KUR fuels: Spent fuel return and reduced enrichment program

    International Nuclear Information System (INIS)

    The Research Reactor Institute of Kyoto University (KURRI) has more than 250 MTR-type HEU spent fuel elements. They have been stored in water pools after irradiation in the Kyoto University Research Reactor (KUR) core. The longest pool residence time is 25 years. In accordance with the Foreign Research Reactor Spent Nuclear Fuel Receipt Program of the United States, sixty KUR spent fuel elements were shipped from KURRI to the Savannah River Site of the US DOE in August 1999. This shipment was done successfully through a public port in Osaka Prefecture, Japan. This is the first shipment in the past twenty-six years after the last shipment through the Yokohama Port. Concerning the use of a public port, we had to solve many issues for public acceptance. In this paper, we describe how we have stored the spent fuels for a long time with high integrity and how we have obtained public acceptance for the transport. So far we have HEU fuels to be used until March 2004, which is already agreed by US DOE. We are looking for candidate LEU fuel materials after HEU, and also spent fuel handling of the new LEU fuel. (author)

  4. Development for analysis system of rods enrichment of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  5. Core configuration of the Syrian reduced enrichment fuel MNSR

    International Nuclear Information System (INIS)

    The possibility of substituting the actual HEU by a LEU or MEU in the Syrian MNSR is investigated through a pre-constructed 3-D detailed model of the reactor. Core configuration does not change if a reduced enrichment fuel (20% u-235, with the same percentages of impurity and eliminating aluminum) is used. The required density for the reactor to be critical in this case would be 7.29 g/cm2. If a specific fuel is used (20 w/o U235, 72 w/o U ), the reactor may not go critical at all. When a MEU fuel is used (45 w/o U235, 40 w/o U), the reactor will restore the same actual initial excess reactivity if 2 standard fuel rods are added to each fuel circle. (author)

  6. Studies of the opportunity to convert the 'ARGUS-90' research reactor with 90% fuel enrichment in U-235 to low-enriched fuel (∼20%)

    International Nuclear Information System (INIS)

    Aiming to assess the consequences of abandoning the employment of highly enriched nuclear fuel (HEU) in the reactor engineering, the opportunity has been studied to convert the 'Argus-90' reactor operating with the uranyl sulphate water solution fuel of 90% enrichment in uranium-235 to low-enriched fuel (LEU) of ∼20% enrichment. A unified technology for the preparation of a solution fuel of 20% and 90% enrichment in U-235 has been confirmed. The effect of low-enriched fuel on the core neutronics parameters has been studied as well as on the efficiency of operating controls of the reactor control and protection system and radiolytic parameters of the solution fuel. (author)

  7. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Science.gov (United States)

    2010-10-01

    ... average fuel economy standard. 525.11 Section 525.11 Transportation Other Regulations Relating to... EXEMPTIONS FROM AVERAGE FUEL ECONOMY STANDARDS § 525.11 Termination of exemption; amendment of alternative average fuel economy standard. (a) Any exemption granted under this part for an affected model year...

  8. 40 CFR 600.510-12 - Calculation of average fuel economy and average carbon-related exhaust emissions.

    Science.gov (United States)

    2010-07-01

    ... must meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy and average carbon-related exhaust emissions. 600.510-12 Section 600.510-12 Protection of...

  9. The development of lower enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U3O8 cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  10. Development of lower enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    As part of the worldwide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt % U alloy and Al-U3O8 cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt % U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behavior of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behavior in high temperature water does not seem much worse. The oxidation and aqueous corrosion of Al-37 wt % U are not much different from those of the Al-U alloys currently used

  11. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  12. 75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

    Science.gov (United States)

    2010-07-29

    ... amended. The introduction of uranium hexafluoride into any module of the National Enrichment Facility is... COMMISSION Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...: Ty Naquin, Project Manager, Uranium Enrichment Branch, Division of Fuel Cycle Safety and...

  13. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U3Si2-Al followed by 0.03% for U3Si-Al, and 0.01% for U3O8-Al fuel. The U3O8-Al fueled reactor gave the maximum ρexcess at BOL which was 21.67% more than the original fuel followed by U3Si-Al which was 2.55% more, while that of U3Si2-Al was 2.50% more than the original UAlx-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U3O8-Al followed by U3Si-Al and then U3Si2-Al fuel.

  14. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U3O8-Al was about 2% more than the original UAlx-Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  15. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  16. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Are there fleet average... PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.55 Are there fleet... that each executive agency meet the fleet average fuel economy standards in place as of January 1...

  17. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR part... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy... Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy §...

  18. ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3

  19. Analysis of the performance of fuel cells BWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles BWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Vargas, S.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: govaj666@hotmail.com

    2008-07-01

    The efficient use of the fuel is one of the objectives in the assemblies design of type BWR. The present tendency in the assemblies design of type BWR is through a radial distribution of enrichments. The present work has like object showing the because of this decision, for what a comparison of the neutronic performance of two fuel cells with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The cells were analyzed with the CASMO-4 code and the obtained results of the behavior of the neutron flow and the power sustain the because of the radial distribution of enrichments. (Author)

  20. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  1. Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle

    International Nuclear Information System (INIS)

    A two-year study of alternatives to the natural uranium fuel cycle in CANDU reactors is summarized. The possible advanced cycles are briefly described. Selection criteria for choosing a cycle for development include resource utilization, economics, ease of implementaton, and social acceptability. It is recommended that a detailed study should be made with a view to the early implementation of the low-enriched uranium cycle. (LL)

  2. RA3: Application of a calculation model for fuel management with SEFE (Slightly Enriched Fuel Elements)

    International Nuclear Information System (INIS)

    The RA-3 (5 MW, MTR) reactor is mainly utilized to produce radioisotopes (Mo-99, I-131, etc.). It started operating with Low Enrichment Uranium (LEU) in 1990, and spends around 12 fuels per year. Although this consumption is small compared to a nuclear power station. It is important to do a good management of them. The present report describes: - A reactor model to perform the Fuel Shuffling. - Results of fuel management simulations for 2 and a half years of operation. Some features of the calculations can be summarized as follows: 1) A 3D calculation model is used with the code PUMA. It does not have experimental adjustments, except for some approximations in the reflector representation and predicts: power, flux distributions and reactivity of the core in an acceptable way. 2) Comparisons have been made with the measurements done in the commissioning with LEU fuels, and it has also been compared with the empirical method (the previous one) which had been used in the former times of operation with LEU fuel. 3) The number of points of the model is approximately 13500, an it can be run in 80386 personal computer. The present method has been verified as a good tool to perform the simulations for the fuel management of RA-3 reactor. It is expected to produce some economic advantages in: - Achieving a better utilization of the fuels. - Leaving more time of operation for radioisotopes production. The activation measurements through the whole core required by the previous method can be significantly reduced. (author)

  3. 76 FR 67765 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding...

    Science.gov (United States)

    2011-11-02

    ... Energy Act of 1954, as amended. The introduction of uranium hexafluoride into any module of the National... COMMISSION Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding... CONTACT: Gregory Chapman, Project Manager, Uranium Enrichment Branch, Division of Fuel Cycle Safety...

  4. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC... Act of 1954, as amended. The introduction of uranium hexafluoride into any module of the National... Regulatory Commission Brian W. Smith, Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety...

  5. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... introduction of uranium hexafluoride (UF 6 ) into cascades numbered 2.9, 2.10, 2.11, 2.12, 3.1, 3.2, 3.3, 3.4..., Uranium Enrichment Branch, Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material...

  6. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration/Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  7. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U3Si2-Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort

  8. Study on the Calculation of Pebble-Bed Reactor Multiplication Factor As a Function of Fuel Kernel Radius at Various Enrichments

    International Nuclear Information System (INIS)

    Main characteristics of PBR comes from utilization of coated particle fuels dispersed in pebble fuels . Because of vibration, fuel kernel can be grouped into cluster and in these cases, neutronic characteristics of pebble fuel significantly changes . In this study, cluster is modeled structural form consisting of uniform cubic cells with eight neighborhood TRISO particles . Neutronic characteristics was investigated by calculating pebble-bed reactor multiplication factor as a function of fuel kernel radius at various enrichments . The calculation results using MCNP5 code with ENDF/BVI neutron library show that keff value depends on the average fuel radius and reaches its minimum when all kernels have the same radius, i.e. 0.0280 cm . With this radius, the total kernel surface area achieves maximum value . The dependence of keff on fuel kernel radius decreases in relation to the increase in uranium enrichment . However, keff value is not affected by fuel kernel radius when the uranium is 100% enriched . From these result, it can be concluded that, exception of uranium enrichment, the selection of fuel kernel radius should be considered thoroughly in designing a PBR, since this parameter provides significant influences on neutronic characteristics of the reactor. (author)

  9. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  10. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  11. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can

  12. Human health impacts avoided by blending highly enriched uranium to low-enriched uranium for commercial nuclear fuel

    International Nuclear Information System (INIS)

    The end of the Cold War and subsequent Strategic Arms Reduction Treaties have resulted in surplus stockpiles of weapons-usable fissile materials in the United States. If not managed properly, these excess stockpiles could pose a danger to national and international security with potential for environmental, safety, and health consequences. The United States has declared 200 tonnes of fissile materials surplus, of which 165 tonnes is highly enriched uranium (HEU). Uranium with 235U enrichments of 20% or greater is considered HEU. The U.S. Department of Energy proposes to blend the surplus HEU to low-enriched uranium (LEU) to eliminate the risk of diversion for nuclear proliferation purposes and, where practical, to reuse the resulting LEU in ways that recover its commercial value. This paper presents the human health risk assessment results for each proposed blending alternative and compares the health impact to that of the commercial nuclear fuel cycle

  13. Performance of low-enriched U3Si2-aluminum dispersion fuel elements in the Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    Six high-density, low-enriched U3Si2-Al dispersion fuel elements have been tested in the Oak Ridge Research Reactor (ORR). The elements were geometrically identical to standard ORR elements. The uranium density in the fuel meat ranged between 4.6 and 5.2 Mg/m3. The elements were fabricated by B and W, CERCA, and NUKEM using their normal materials and fabrication practices, with minor modifications necessitated by the new fuel. The U3Si2 contained minor amounts of USi, U3Si and/or uranium solid solution. The elements were irradiated to approximately normal ORR burnup, and three elements were irradiated twice as long, to average burnups of ∼80% of the initially contained 235U, well above the burnups normally achieved in research and test reactors. Peak burnups of 98% were achieved. Following suitable cooling periods, the elements were subjected to a series of nondestructive and destructive examinations. The behavior of the fuel was found to be entirely consistent with the known irradiation behavior of the constituent phases. The extremely stable swelling behavior of the U3Si2 phase dominated in all cases. The plates showed small, uniform thickness changes. Blister threshold temperatures were ≥5500C. It is concluded that low-enriched U3Si2-Al dispersion fuel elements will perform at least as well in research and test reactors with power densities up to that of the ORR as the highly enriched UAl/sub x/-Al and U3O8-Al dispersion fuels currently being used. There were no indications that use of this fuel under substantially more stringent conditions might be precluded. 10 refs., 87 figs., 9 tabs

  14. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    Energy Technology Data Exchange (ETDEWEB)

    Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  15. Simulation studies of diesel engine performance with oxygen enriched air and water emulsified fuels

    Energy Technology Data Exchange (ETDEWEB)

    Assanis, D.N.; Baker, D. (Illinois Univ., Urbana, IL (USA)); Sekar, R.R.; Siambekos, C.T.; Cole, R.L.; Marciniak, T.J. (Argonne National Lab., IL (USA))

    1990-01-01

    A computer simulation code of a turbocharged, turbocompound diesel engine was modified to study the effects of using oxygen-enriched combustion air and water-emulsified diesel fuels. Oxygen levels of 21 percent to 40 percent by volume in the combustion air were studied. Water content in the fuel was varied from 0 percent to 50 percent mass. Simulation studies and a review and analysis of previous work in this area led to the following conclusions about expected engine performance and emissions: the power density of the engine is significantly increased by oxygen enrichment. Ignition delay and particulate emissions are reduced. Combustion temperatures and No{sub x} emissions are increased with oxygen enrichment but could be brought back to the base levels by introducing water in the fuel. The peak cylinder pressure which increases with the power output level might result in mechanical problems with engine components. Oxygen enrichment also provides an opportunity to use cheaper fuel such as No. 6 diesel fuel. Overall, the adverse effects of oxygen enrichment could be countered by the addition of water and it appears that an optimum combination of water content, oxygen level, and base diesel fuel quality may exist. This could yield improved performance and emissions characteristics compared to a state-of-the-art diesel engine. 9 refs., 8 figs.

  16. Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-05-15

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  17. Conversion of highly enriched fuel of the research reactors in the framework of international initiatives

    International Nuclear Information System (INIS)

    U.S. Department of Energy and the National Nuclear Security Administration launched an initiative to reduce the risk of theft and illegal use of nuclear and radioactive materials. In the framework of this initiative in the performance of the Russian fuel return program from the research reactors(RRRFR) the staff and experts made the restitution of highly enriched fuel to Russia and performed the conversion of the INRNAS of Ukraine research reactor to the fuel with low-enriched uranium (LEU < 20 % U-235). The works were also carried out on the systems modernization which are important to the safe operation of the reactor. Ukraine has fulfilled its international obligations and released the territory from the highly enriched uranium in time

  18. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2002-01-01

    Full Text Available This paper presents the relevant data related to the recent shipment (August 2002 of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  19. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U3O8-Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  20. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  1. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Science.gov (United States)

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N

    2012-10-01

    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

  2. Plutonium-enriched thermal fuel production experience in Belgium

    International Nuclear Information System (INIS)

    Taking into account the strategic aspects of nuclear energy such as availability and sufficiency of resources and independence of energy supply, most countries planning to use plutonium look mainly to its use in fast reactors. However, by recycling the recovered uranium and plutonium in light water reactors, the saving of the uranium that would otherwise be required could already be higher than 35%. Therefore, until fast reactors are introduced, for macro- or microeconomic reasons, the plutonium recycle option seems to be quite valuable for countries having the plutonium technology. In Belgium, Belgonucleaire has been developing the plutonium technology for more than 20 yr and has operated a mixed oxide fuel fabrication plant since 1973. The past ten years of plant operation have provided for many improvements and relevant new documented experiences establishing a basis for new modifications that will be beneficial to the intrinsic quality, overall safety, and economy of the fuel

  3. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  4. Enrichment effects on CANDU-SEU spent fuel Monte Carlo shielding analysis

    International Nuclear Information System (INIS)

    Shielding analyses are an essential component of the nuclear safety, the estimations of radiation doses in order to reduce them under specified limitation values being the main task here. According to IAEA data, more than 10 millions packages containing radioactive materials are annually transported world wide. All the problems arisen from the safe radioactive materials transport assurance must be carefully settled. Last decade, both for operating reactors and future reactor projects, a general trend to raise the discharge fuel burnup has been recorded world wide. For CANDU type reactors, the most attractive solution seems to be SEU and RU fuels utilization. The basic tasks accomplished by the shielding calculations in a nuclear safety analysis consist in dose rates calculation, to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper aims to study the effects induced by fuel enrichment variation on CANDU-SEU spent fuel photon dose rates for a Monte Carlo shielding analysis applied to spent fuel transport after a defined cooling period in the NPP pools. The fuel bundles projects considered here have 43 Zircaloy rods, filled with SEU fuel pellets, the fuel having different enrichment in U-235. All the geometrical and material data related on the cask were considered according to the shipping cask type B model. After a photon source profile calculation by using ORIGEN-S code, in order to perform the shielding calculations, Monte Carlo MORSE-SGC code has been used, both codes being included in the ORNL's SCALE 5 system. The photon dose rates to the shipping cask wall and in air, at different distances from the cask, have been estimated. Finally, a photon dose rates comparison for different fuel enrichments has been performed. (author)

  5. Study of Reduced-Enrichment Uranium Fuel Possibility for Research Reactors

    Directory of Open Access Journals (Sweden)

    Ruppel V.A.

    2015-01-01

    Full Text Available Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5% for UO2 fuel and by 7% for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4% for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1% less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

  6. High-density reduced-enrichment fuels for Research and Test Reactors

    International Nuclear Information System (INIS)

    Development and irradiation testing of high-density fuels have been conducted by the US RERTR Program in order to provide the technical means to reduce the enrichment of fuels for research and test reactors. The traditional aluminum dispersion fuel technology has been extended to include the highest practical loadings of uranium-aluminide (UAl/sub x/, 2.3 MgU/m3), uranium-oxide (U3O8, 3.2 MgU/m3), and uranium-silicide (U3Si2, 5.5 MgU/m3; U3Si, 7.0 MgU/m3) fuels. A third uranium-silicide alloy, U3SiAl (U + 3.5 wt % Si + 1.5 wt % Al) has been found to perform poorly at high burnup. Testing of miniature fuel plates and full-sized fuel elements is at an advanced stage for the highest loadings of the aluminide and oxide fuels and intermediate loadings of the silicide fuels, and good results have been obtained for low-enriched uranium. The data obtained to date are discussed. 1 reference, 3 figures, 1 table

  7. Parametric evaluation of mixed (low and high enriched) fuel core for a swimming pool type research reactor

    International Nuclear Information System (INIS)

    A study has been carried out to evaluate the performance of a swimming pool type research reactor core comprised of mixed (low and high enriched) uranium fuel. The study includes the calculations of core reactivity, worth of control rods and core criticality at the Beginning Of Life (BOL) of the core and for two operating conditions Cold Zero Power (CZP) and Hot Full Power (HFP). Further, to ensure safe and stable operation of the core from nuclear design point of view, average power densities in the fuel region, power peaking factors, axial power distribution in the hot channel and reactivity feed back coefficients have also been calculated. Two group fluxes have also been determined at different irradiation locations. All these calculations have been carried out employing reactor lattice code WIMS-D14 and reactor analysis code CITATION The calculated results show reasonably good agreement with the quoted operational data of the previous LEU cores. (Author)

  8. Kinetic study of the Tehran research reactor core with low enriched fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pazirandeh, A.; Afshar Bakeshloo, A. [Tehran Univ. (Iran, Islamic Republic of). Physics Dept.; Bartsch, G. [Technische Univ. Berlin (Germany). Inst. fuer Energietechnik

    1997-11-01

    For evaluating the performance of the newly refuelled Tehran Research Reactor core with low enriched uranium fuel (LEU) in transient states a two group time dependent diffusion equation code (COSTANZA) was used. This paper presents results of calculations of the fast transients, revealing the steady performance of the core and fuel integrity during transient for a probable reactivity insertion of less than or equal dollar 1.5/0.5 s. The temperature dependant reactivity coefficients of the Doppler resonance broadening effect and of the moderator absorption cross section change and density dilution were calculated using cell-averaged 69 energy group WIMS-D/4 for two main libraries, old library and WIMKAL88, to 13 groups. The two group parameters for the COSTANZA code were also obtained by WIMS-D/4. (orig.) [Deutsch] Zur Bewertung der Leistungsfaehigkeit des neu beladenen Teheraner Forschungsreaktors mit niedrig angereichertem Uranbrennstoff bei Reaktivitaetstransienten wurde ein 2-Gruppen zeitabhaengiges Diffusionsprogramm COSTANZA verwendet. In der vorliegenden Arbeit werden Ergebnisse der Berechnung schneller Transienten vorgestellt, die das Verhalten des Reaktorkerns bzw. die Integritaet der Brennstaebe waehrend der Transienten fuer eine Reaktivitaetsaenderung von kleiner oder gleich Dollar 1.5/0.5 s zeigen. Die temperaturabhaengigen Reaktivitaetskoeffizienten der Doppler-Verbreitung im Brennstoff sowie der Dichteaenderung und der Neutronenabsorption im Moderator wurden mit Hilfe zellengemittelter 69 Energie-Gruppen der Datenbank WIMS-D/4 und fuer 13 Energiegruppen mit der Datenbank WIMKAL 88 ermittelt. Die Zweigruppendaten fuer das COSTANZA-Programm wurden ebenfalls mit Hilfe von WIMS-D/4 bestimmt. (orig.)

  9. Research and Test Reactor Conversion to Low Enriched Uranium Fuel: Technical and Programmatic Progress

    International Nuclear Information System (INIS)

    The U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of High Enriched Uranium (HEU) fuel in research reactors by converting them to low enriched uranium (LEU) fuel. In 2004, the reactor conversion program became the driving pilar of the Global Threat Reduction Initiative (GTRI), a program established by the U.S. DOE's National Nuclear Security Administration. The overall GTRI objectives are the conversion, removal or protection of vunerable civilian radiological and nuclear material. As part of the GTRI, the Conversion Program has accelerated the schedules and plans for conversion of additional research reactors operating with HEU. This paper provides an update on the progress made since 2007 and describes current technical challenges that the program faces. (author)

  10. The Shipment of Russian-Origin Highly Enriched Uranium Research Reactor Spent Nuclear Fuel from Belarus

    Energy Technology Data Exchange (ETDEWEB)

    Sikorin, S.N.; Polazau, S.A.; Hryharovich, T.K. [Joint Institute for Power and Nuclear Research ' Sosny' , Academik Krasin Street, Minsk (Belarus); Bolshinsky, I. [Idaho National Laboratory, N. Fremont Avenue Idaho Falls, Idaho (United States); Thomas, J.E. [Savannah River National Laboratory, Aiken, South Carolina (United States)

    2011-07-01

    In October 2010, the Global Threat Reduction Initiative and the Joint Institute for Power and Nuclear Research - 'Sosny' of the National Academy of Sciences of the Republic of Belarus completed a shipment that returned 43 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel to the Russian Federation. The spent fuel was legacy material, discharged from the two decommissioned reactors, the Pamir-630D mobile reactor and the IRT-M research reactor. This shipment marked the complete removal of all HEU spent nuclear fuel from Belarus. This paper discusses the planning, preparations, and coordination required to complete this important international shipment successfully. (author)

  11. The Shipment of Russian-Origin Highly Enriched Uranium Research Reactor Spent Nuclear Fuel from Belarus

    International Nuclear Information System (INIS)

    In October 2010, the Global Threat Reduction Initiative and the Joint Institute for Power and Nuclear Research - 'Sosny' of the National Academy of Sciences of the Republic of Belarus completed a shipment that returned 43 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel to the Russian Federation. The spent fuel was legacy material, discharged from the two decommissioned reactors, the Pamir-630D mobile reactor and the IRT-M research reactor. This shipment marked the complete removal of all HEU spent nuclear fuel from Belarus. This paper discusses the planning, preparations, and coordination required to complete this important international shipment successfully. (author)

  12. Critical experiments on minimal-content gadolinia for above-5wt% enrichment fuels in Toshiba NCA

    International Nuclear Information System (INIS)

    A concept of 'minimal-content gadolinia' with a content of less than several hundred ppm mixed in the 'above-5wt% enrichment UO2 fuel' for super high burnup is proposed for ensuring the criticality safety in the UO2 fuel fabrication facility for light water reactors (LWRs) without increase in investment cost. Required gadolinia contents calculated were from 53 to 305 ppm for enrichments of UO2 powders for boiling water reactor (BWR) fuel from 6 to 10 wt%. It is expected that the minimal-content gadolinia yields an acceptable reactivity suppression at the beginning of operating cycle and no reactivity penalty at the end of operating cycle due to no residual gadolinium. A series of critical experiments were carried out in the Toshiba Nuclear Critical Assembly (NCA). Reactivity effects of the gadolinia were measured to clarify the nuclear characteristics, and the measured values and the calculated values agreed within 5%. (author)

  13. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    Energy Technology Data Exchange (ETDEWEB)

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  14. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  15. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

  16. Conversion and Evaluation of the University of Massachusetts Lowell Research Reactor From High-Enriched To Low-Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    The process for converting the University of Massachusetts Lowell Research Reactor (UMLRR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel began in 1988. Several years of design reviews, computational modeling, and thermal hydraulic analyses resulted in a preliminary reference core design and configuration based on 20 standard, MTR-type, flat-plate, 19.75% enriched, uranium silicide (u3Si2) fuel elements. A final safety analysis for the fuel conversion was submitted to the Nuclear Regulatory Commission (NRC) in 1993. The NRC made two additional requests for additional information and supplements were submitted in 1994 and 1997. The new UMLRR Reactor Supervisor initiated an effort to change the LEU reference core configuration to eliminate a complicated control rod modification needed for the smaller core

  17. Argentine activities related to the development of low enriched fuel elements

    International Nuclear Information System (INIS)

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the USA and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project) has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown. The densities reached were within the range of 3.12 to 3.58 g/cm3 for U3O8-Al, 2.99 to 3.09 g/cm3 for UAl2-Al and 5.18 to 6.10 g/cm3 for U3Si-Al. If further miniplates can be irradiated, it is the purpose of the program to research uranium densities of 3.5 g/cm3 in UAl2-Al and 6.5 g/cm3 in U3Si-Al

  18. 77 FR 64051 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Science.gov (United States)

    2012-10-18

    ... which was published in the Federal Register of Monday, October 15, 2012 (77 FR 62624). The final rule... Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy...

  19. Atomics International fuel fabrication facility and low enrichment program. Part 2

    International Nuclear Information System (INIS)

    Most of you know our company from the last meeting in May in Vienna, so I won't steal your time with explaining and demonstrating the same techniques that we have heard this morning f rom the other speakers. I would just take some words to explain the order of business with highly enriched uranium. NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 2.5 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In TABLE I are the different fuel types (see column 1) and then we have the fabrication in column 2; (The reason that I use the blackboard this morning is that I try to demonstrate all the techniques. However, all the speakers before me did this and in theory we are not so far away from each other.) the experience of my company in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7, and column 8 is the estimated cost of 6 and 7. There is just one fuel that is not in this summary and that is U-Zr

  20. Radiological health aspects of commercial uranium conversion, enrichment, and fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Stoetzel, G.A.; Hoenes, G.R.; Cummings, F.M.; McCormack, W.D.

    1982-11-01

    Detailed information concerning occupational exposures, health physics practices, and regulatory procedures at commercial conversion, enrichment and fuel fabrication facilities is given. Sites visits were the primary source of information, which is divided into four sections. The first section discusses health physics practices that are common to the conversion, enrichment, and fuel fabrication phases of the commercial uranium industry. The next three sections review process descriptions, radiological health practices, and regulatory procedures for the three phases. Nonradiological exposures are considered only as they influence the interpretation of the health effects of radiological exposures. The review of regulatory procedures indicates the types of exposure evaluation records being kept on uranium workers and the responsibility for maintaining the records.

  1. Radiological health aspects of commercial uranium conversion, enrichment, and fuel fabrication

    International Nuclear Information System (INIS)

    Detailed information concerning occupational exposures, health physics practices, and regulatory procedures at commercial conversion, enrichment and fuel fabrication facilities is given. Sites visits were the primary source of information, which is divided into four sections. The first section discusses health physics practices that are common to the conversion, enrichment, and fuel fabrication phases of the commercial uranium industry. The next three sections review process descriptions, radiological health practices, and regulatory procedures for the three phases. Nonradiological exposures are considered only as they influence the interpretation of the health effects of radiological exposures. The review of regulatory procedures indicates the types of exposure evaluation records being kept on uranium workers and the responsibility for maintaining the records

  2. Atomics international fuel fabrication facility and low enrichment program [contributed by H.W. Hassel, NUKEM

    International Nuclear Information System (INIS)

    NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 25 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would-just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In the table are the different fuel types (see column) and then we have the fabrication in column 2 the experience of my comp any in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7 and column is the estimated cost of 6 and 7 There is just one fuel that is not in this summary and that is U-Zr. We now see how complex and sophisticated this business is. I have told you already that we have installed for a lot of millions of Deutsche Mark the physical protection, storage vaults and things like that. Now we have to investigate all these different types of fuels for, as you see, a lot of money. Maybe these are a lot of optimistic figures; anyway the question is, does this make all the overall nuclear situation worldwide easier or not. One cannot answer for the moment, but anyway we have a lot of problems

  3. 75 FR 80430 - Passenger Car and Light Truck Average Fuel Economy Standards Request for Product Plan Information...

    Science.gov (United States)

    2010-12-22

    ... Average Fuel Economy Standards Request for Product Plan Information--Model Years 2010-2025 AGENCY... Intent, 75 FR 62739 (Oct. 13, 2010). \\5\\ Available at http://www.nhtsa.gov/fuel-economy (last accessed... for comments. SUMMARY: The purpose of this request for comments is to acquire updated...

  4. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  5. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  6. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  7. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    Energy Technology Data Exchange (ETDEWEB)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  8. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    Energy Technology Data Exchange (ETDEWEB)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  9. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO2 and PuO2-UO2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO2 rods at two enrichments (2.35 wt % and 4.31 wt % 235U) and on mixed fuel-water assemblies of UO2 and PuO2-UO2 rods containing 4.31 wt % 235U and 2 wt % PuO2 in natural UO2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  10. Comparison of heuristic optimization techniques for the enrichment and gadolinia distribution in BWR fuel lattices and decision analysis

    International Nuclear Information System (INIS)

    Highlights: • Different metaheuristic optimization techniques were compared. • The optimal enrichment and gadolinia distribution in a BWR fuel lattice was studied. • A decision making tool based on the Position Vector of Minimum Regret was applied. • Similar results were found for the different optimization techniques. - Abstract: In the present study a comparison of the performance of five heuristic techniques for optimization of combinatorial problems is shown. The techniques are: Ant Colony System, Artificial Neural Networks, Genetic Algorithms, Greedy Search and a hybrid of Path Relinking and Scatter Search. They were applied to obtain an “optimal” enrichment and gadolinia distribution in a fuel lattice of a boiling water reactor. All techniques used the same objective function for qualifying the different distributions created during the optimization process as well as the same initial conditions and restrictions. The parameters included in the objective function are the k-infinite multiplication factor, the maximum local power peaking factor, the average enrichment and the average gadolinia concentration of the lattice. The CASMO-4 code was used to obtain the neutronic parameters. The criteria for qualifying the optimization techniques include also the evaluation of the best lattice with burnup and the number of evaluations of the objective function needed to obtain the best solution. In conclusion all techniques obtain similar results, but there are methods that found better solutions faster than others. A decision analysis tool based on the Position Vector of Minimum Regret was applied to aggregate the criteria in order to rank the solutions according to three functions: neutronic grade at 0 burnup, neutronic grade with burnup and global cost which aggregates the computing time in the decision. According to the results Greedy Search found the best lattice in terms of the neutronic grade at 0 burnup and also with burnup. However, Greedy Search is

  11. Hydrogen enriched compressed natural gas (HCNG: A futuristic fuel for internal combustion engines

    Directory of Open Access Journals (Sweden)

    Nanthagopal Kasianantham

    2011-01-01

    Full Text Available Air pollution is fast becoming a serious global problem with increasing population and its subsequent demands. This has resulted in increased usage of hydrogen as fuel for internal combustion engines. Hydrogen resources are vast and it is considered as one of the most promising fuel for automotive sector. As the required hydrogen infrastructure and refueling stations are not meeting the demand, widespread introduction of hydrogen vehicles is not possible in the near future. One of the solutions for this hurdle is to blend hydrogen with methane. Such types of blends take benefit of the unique combustion properties of hydrogen and at the same time reduce the demand for pure hydrogen. Enriching natural gas with hydrogen could be a potential alternative to common hydrocarbon fuels for internal combustion engine applications. Many researchers are working on this for the last few years and work is now focused on how to use this kind of fuel to its maximum extent. This technical note is an assessment of HCNG usage in case of internal combustion engines. Several examples and their salient features have been discussed. Finally, overall effects of hydrogen addition on an engine fueled with HCNG under various conditions are illustrated. In addition, the scope and challenges being faced in this area of research are clearly described.

  12. Computational design of parameters of IRT-2M fuel with enrichment below 20% for low power research reactors

    International Nuclear Information System (INIS)

    This article focuses briefly on characteristics of a possible procedure during reduction of fuel enrichment of two research reactors in the Czech Republic, i.e., LVR-15 research reactor (power up to 15 MW) at NRI Rez and VR-1 training reactor (power up to 5 kW) at CTU Prague. Both reactors are now operating with fuel enriched to 36% of 235U. While the LVR-15 reactor uses Russian IRT-2M fuel, the VR-1 reactor has been operating on IRT-3M fuel for five years already. The goal for both reactors until now was to use Russian IRT-4M fuel with 235U enrichment below 20%. The original idea that the LVR-15 reactor would go through the IRT-3M fuel during the transition to IRT-4M fuel now seems baseless. The article hence shows a possible solution to the current situation for the VR-1 reactor. A convenient solution (based on consultations with the Russian producer) could be a preparation of fuel of IRT-2M geometry with enrichment to 20% of 235U. Such a fuel would not be intended for power research reactors in the first step but for reactors with power up to 100 - 200 kW. The article presents a proposal of this fuel (said proposal was created on the basis of many years' standing experience of operation at VR-1 reactor) and verifying calculations for selected configurations. As for enrichment, matrix, and content of uranium the proposal is based on verified capability of the Russian producer. Emphasis is placed on the necessity of the fuel having a long lifetime in the light water reactors. (author)

  13. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  14. Bulgarian experience with the implementation of {sup 235}U enriched fuel in WWER-1000 units

    Energy Technology Data Exchange (ETDEWEB)

    Dobrevski, Ivan; Zaharieva, Neili [Bulgarian Academy of Sciences, Sofia (Bulgaria). Inst. for Nuclear Research and Nuclear Energy

    2009-12-15

    This paper reports on the results of the implementation of TVSA fuel assemblies with up to 4.3 % {sup 235}U enrichment and an integrated burnable absorber (Gd) (U-Gd{sub 2}O{sub 3} fuel with 5 % Gd{sub 2}O{sub 3}) in WWER-1000 reactors at Kozloduy Nuclear Power Plant in Bulgaria. Data from the first cycle with 100 % TVSA assemblies show that plant staff was able to maintain the coolant water chemistry within the range demanded by the plant's primary circuit water chemistry requirements. Data indicate that the corrosion processes in the primary circuit remained on the same low level as during previous cycles. (orig.)

  15. Calculation of parameters for inspection planning and evaluation: low enriched uranium conversion and fuel fabrication facilities

    International Nuclear Information System (INIS)

    As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities at low-enriched uranium (LEU) conversion and fuel fabrication facilities. This report presents the results and conclusions of those analyses. Implementation of IAEA safeguards at LEU conversion and fuel fabrication facilities must take into account a variety of practical problems and constraints. One of the key concerns is the problem of flow verification, especially product verification. The objective of this report is to help put the problem of flow verification in perspective by presenting the results of some specific calculations of inspection effort and probability of detection for various product measurement strategies. In order to provide quantitative information about the advantages and disadvantages of the various strategies, eight specific cases were examined

  16. Irradiation performance of reduced-enrichment fuels tested under the U.S. RERTR Program

    International Nuclear Information System (INIS)

    Considerable progress in the irradiation testing of high-density, reduced-enrichment fuels has been made during the past year. Miniplates containing UAI, U3Si2, U3Si1.5, U3Si, U3SiCu, and U6Fe have been Irradiated. Postirradiation examinations have revealed that breakaway swelling has occurred in 6.4-Mg U/m3 U3Si plates at ∼2.8 x 1027 fissions/m3 and in U6Fe plates at ∼1.4 x 1027 fissions/m3. U3Si2 plates continue to perform satisfactorily. The testing of full-sized fuel elements in the ORR and the SILOE reactor have continued with good results. Postirradiation examinations are confirming the satisfactory performance of these elements. (author)

  17. Non-destructive evaluation methods to improve quality control in low enrichment MTR fuel plate production

    International Nuclear Information System (INIS)

    This paper summarises some preliminary non-destructive measurements made recently at the Harwell Laboratory on a prototype low enrichment MTR fuel plate. The measurements were intended to indicate the potential of two different techniques for improving quality control in plate production. Pulse Video Thermography (PVT) is being considered as an alternative to ultrasound transmission measurements for the detection and sizing of lack of thermal bonding between the fuel and the clad layers, either to verify the indications from the established ultrasonic methods before destroying the plate or as a replacement method of inspection. High frequency pulse-echo ultrasonics is being considered for providing maps of clad layer thickness on each side of the plate. The measurements have indicated the potential for both methods, but more work is required, using a test plate containing controlled defects, to establish their capability. (orig.)

  18. Optimization of enrichment distributions in nuclear fuel assemblies loaded with uranium and plutonium via a modified linear programming technique

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed OXide (MOX) fuels. MOX isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide-range of applicability of the optimization technique. The features of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or

  19. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  20. Postirradiation examination of a low enriched U3Si2-Al fuel element manufactured and irradiated at Batan, Indonesia

    International Nuclear Information System (INIS)

    The first low-enriched U3Si2-Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U3Si2-Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

  1. 76 FR 26996 - Notice of Intent To Prepare an Environmental Impact Statement for New Corporate Average Fuel...

    Science.gov (United States)

    2011-05-10

    ... (Sept. 2010) \\17\\ See 75 FR 62739 (Oct. 13, 2010). In response to the President's call to provide... Regulations, 46 FR 18026 (1981) (emphasis added). Alternatives calculated at the upper point and at the lower... Environmental Impact Statement for New Corporate Average Fuel Economy Standards AGENCY: National Highway...

  2. Argentine activities related to the development of low enriched fuel elements

    International Nuclear Information System (INIS)

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the U.S.A. and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project), has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown in Table 1. At a meeting held in Vienna in March, 1980, the CNEA stated that its development of fuels with low enrichment would be in two fuel lines: UAlx-Al and U3O8-Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm3 for the UAIx-Al line and 2.4-3.0 g/cm3 for the U3O8 line. At the international meeting held at ANL in November, 1980, and after having received depleted uranium and uranium with 20% and 45% enrichment (purchased from the U.S.A. for manufacturing miniplates and possible standard fuels) to carry on the proposed development, CNEA anticipated -- after its first tests -- that the conditions were satisfactory for reaching uranium densities of 2.4-3.0 g/cm3 in U3O8-Al fuel and of 2.4 g/cm3 in UAIx-Al fuel. In February 1981, after Argentina accepted the obligation of paying for the irradiation service, authorization was obtained for irradiating miniplates in the Oak Ridge Reactor within the RERTR Program. In June 1981, the first set of miniplates was sent to Oak Ridge National Laboratory (ORNL). The maximum actual densities reached at that time were 3.12 g/cm3 with U3O8-Al and 2.52 g/cm3 with UAlx-Al. During a visit of the CNEA Project Technical Manager to the Argonne National Laboratory (ANL) in July 1981, and after exchanging ideas with ANL professional staff, the CNEA decided to incorporate a new line of development, that of U3Si-Al. Three months later, in October 1981 a second set of

  3. Evaluating the effect of using different sets of enrichment for FAs on fuel management optimization using CA

    International Nuclear Information System (INIS)

    In nuclear reactor core design, achieving the optimized arrangement of fuel assemblies (FAs) is the most important step towards satisfying safety and economic requirements. In most studies, nuclear fuel optimizations have been performed by using a finite number of different types of FAs. However the effect of FA numbers with different enrichments and the difference between their maximum and minimum enrichment values can be important and should be evaluated in the optimization process. This research is aimed at evaluating the effect of using different enrichment values for FAs. This issue has been investigated by focusing on two parameters, namely, the initially selected enrichment and the difference between the minimum and maximum enrichments applied in the core design. In the previous studies of nuclear fuel management, these parameters have been kept as fixed quantities and considered as initial assumptions in the optimization process. Therefore, to achieve an optimized arrangement of the core, the proper values of these parameters have to be determined. For this purpose a parameter (δ) served through the optimization process to show the effect of the difference between the enrichment values of FAs. Another parameter named ε0 shows the minimum enrichment of FAs. These parameters are defined based on a factor named Fuel Quality Factor (FQF) as a characteristic of fuel composition. FQF is shown by Z(r) is also used through the optimization process for achieving the smooth distribution of power. The values of Z(r) are calculated using the MCNP code. This methodology is applied to a VVER-1000 nuclear reactor core in order to minimize the local power peaking factor (Pq). For finding the best configuration of FAs in the core, Cellular Automata (CA) is applied as a powerful and reliable tool. The computer codes WIMS and CITATION are used for core calculations. The results provide a comprehensive view of VVER-1000 reactor core configuration for different groups of

  4. Thermal-hydraulic analysis for core conversion to the use of low-enriched uranium fuels in the KUR

    International Nuclear Information System (INIS)

    A feasibility study has been performed on the core conversion to the use of low-enriched uranium (LEU) fuels in the KUR. Five fuel element geometries are studied. For each fuel element, the relation between the pressure drop and the flow rate, critical heat flux, and heat fluxes for the onset of flow instability and the onset of nucleate boiling are calculated using the computer code PLTEMP3.MOD1 which has been developed for this analysis. The effect of fuel material (UAL/sub x/-Al, U3O8-Al and U3Si2-Al) on the peak fuel temperatures is also studied. A particular interest in the mixed core which may be constructed on the way to the use of LEU fuels, the change in the bypass flow rate due to the change in the gap between different fuel elements is investigated

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    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.

  6. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  7. Recovery of enriched uranium from waste solution obtained from fuel manufacture laboratories

    International Nuclear Information System (INIS)

    Reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of microgram to gram quantities of Uranium (19.7% enriched with 235U) from highly impure solution. The processing of Uranium compounds for atomic energy project especially in FMPP (fuel manufacture pilot plant) gives rise to a variety of wastes in which the Uranium content is of considerable importance. The recovery of Uranium from concentrated mother liquors produced from ADU (ammonium diuranate) precipitation, as well as those due to ADU washing is studied in this work. Column of Poly-trifluoro-monochloro-ethylene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains Uranium. Impurities are eluted with 6.5 M HCl, and the Uranium is eluted with water and the recovery of Uranium is better than 94%. (author)

  8. Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO2 required. (Author)

  9. Test Operation of Oxygen-Enriched Incinerator for Wastes From Nuclear Fuel Fabrication Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.-G.; Yang, H.cC.; Park, G.-I.; Kim, I.-T.; Kim, J.-K.

    2002-02-26

    The oxygen-enriched combustion concept, which can minimize off-gas production, has been applied to the incineration of combustible uranium-containing wastes from a nuclear fuel fabrication facility. A simulation for oxygen combustion shows the off-gas production can be reduced by a factor of 6.7 theoretically, compared with conventional air combustion. The laboratory-scale oxygen enriched incineration (OEI) process with a thermal capacity of 350 MJ/h is composed of an oxygen feeding and control system, a combustion chamber, a quencher, a ceramic filter, an induced draft fan, a condenser, a stack, an off-gas recycle path, and a measurement and control system. Test burning with cleaning paper and office paper in this OEI process shows that the thermal capacity is about 320 MJ/h, 90 % of design value and the off-gas reduces by a factor of 3.5, compared with air combustion. The CO concentration for oxygen combustion is lower than that of air combustion, while the O2 concentration in off-gas is kept above 25 vol % for a simple incineration process without any grate. The NOx concentration in an off-gas stream does not reduce significantly due to air incoming by leakage, and the volume and weight reduction factors are not changed significantly, which suggests a need for an improvement in sealing.

  10. Surface strontium enrichment on highly active perovskites for oxygen electrocatalysis in solid oxide fuel cells

    KAUST Repository

    Crumlin, Ethan J.

    2012-01-01

    Perovskite oxides have high catalytic activities for oxygen electrocatalysis competitive to platinum at elevated temperatures. However, little is known about the oxide surface chemistry that influences the activity near ambient oxygen partial pressures, which hampers the design of highly active catalysts for many clean-energy technologies such as solid oxide fuel cells. Using in situ synchrotron-based, ambient pressure X-ray photoelectron spectroscopy to study the surface chemistry changes, we show that the coverage of surface secondary phases on a (001)-oriented La 0.8Sr 0.2CoO 3-δ (LSC) film becomes smaller than that on an LSC powder pellet at elevated temperatures. In addition, strontium (Sr) in the perovskite structure enriches towards the film surface in contrast to the pellet having no detectable changes with increasing temperature. We propose that the ability to reduce surface secondary phases and develop Sr-enriched perovskite surfaces of the LSC film contributes to its enhanced activity for O 2 electrocatalysis relative to LSC powder-based electrodes. © 2012 The Royal Society of Chemistry.

  11. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  12. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    International Nuclear Information System (INIS)

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (20 fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 1020 fissions, the number of fissions equals about 1028, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada

  13. Conceptual design and economic analysis of a light water reactor fuel enricher/regenerator. FY 1978 year-end report

    Energy Technology Data Exchange (ETDEWEB)

    Grand, P; Kouts, H J; Powell, J R; Steinberg, M; Takahashi, H

    1979-05-01

    A study has been performed to evaluate the use of high-energy particle accelerators as nuclear fuel enrichers and nuclear fuel regenerators. This builds on ideas that have been current for many years. The new study has, however, explored some novel approaches that have not been examined before. A specific conceptual system chosen for more detailed study would stretch the energy available from natural uranium by a factor of about 3, reduce the separative work requirements by a factor of about 4, and reduce the volume of spent fuel to be stored by a factor of 2, compared to the current once-through light water reactor (LWR) fuel cycle. The concept avoids the need for chemical reprocessing of spent fuel, and would permit continued use of LWR's beyond the time when limitations on fuel resources might otherwise lead to their being phased out. This concept, which is called the Linear Accelerator Fuel Enricher/Regenerator, is therefore viewed as offering a practical means of stretching the use of the nuclear fuel resource in the framework of the existing light water reactor fuel cycle. This report describes and analyzes the concept referred to. An explanation of the principles underlying the concept is given. Particular attention is devoted to engineering feasibility, proliferation resistance, and economics. It is seen that the concept draws on only proven technology as regards bothaccelerator design and the fuel irradiation process, and is adapted to existing LWR designs with no change except in fuel-handling practices. A preliminary evaluation of radiation damage, coolant options, and power conversion systems is provided. Neutronic, thermal-hydraulic, and burnup calculations are presented. An analysis is made of fuel economy. Approximate costs of electric power produced using this concept are evaluated and discussed. Estimated development costs of commercialization are provided.

  14. Enrichment of microbial electrolysis cell biocathodes from sediment microbial fuel cell bioanodes.

    Science.gov (United States)

    Pisciotta, John M; Zaybak, Zehra; Call, Douglas F; Nam, Joo-Youn; Logan, Bruce E

    2012-08-01

    Electron-accepting (electrotrophic) biocathodes were produced by first enriching graphite fiber brush electrodes as the anodes in sediment-type microbial fuel cells (sMFCs) using two different marine sediments and then electrically inverting the anodes to function as cathodes in two-chamber bioelectrochemical systems (BESs). Electron consumption occurred at set potentials of -439 mV and -539 mV (versus the potential of a standard hydrogen electrode) but not at -339 mV in minimal media lacking organic sources of energy. Results at these different potentials were consistent with separate linear sweep voltammetry (LSV) scans that indicated enhanced activity (current consumption) below only ca. -400 mV. MFC bioanodes not originally acclimated at a set potential produced electron-accepting (electrotrophic) biocathodes, but bioanodes operated at a set potential (+11 mV) did not. CO(2) was removed from cathode headspace, indicating that the electrotrophic biocathodes were autotrophic. Hydrogen gas generation, followed by loss of hydrogen gas and methane production in one sample, suggested hydrogenotrophic methanogenesis. There was abundant microbial growth in the biocathode chamber, as evidenced by an increase in turbidity and the presence of microorganisms on the cathode surface. Clone library analysis of 16S rRNA genes indicated prominent sequences most similar to those of Eubacterium limosum (Butyribacterium methylotrophicum), Desulfovibrio sp. A2, Rhodococcus opacus, and Gemmata obscuriglobus. Transfer of the suspension to sterile cathodes made of graphite plates, carbon rods, or carbon brushes in new BESs resulted in enhanced current after 4 days, demonstrating growth by these microbial communities on a variety of cathode substrates. This report provides a simple and effective method for enriching autotrophic electrotrophs by the use of sMFCs without the need for set potentials, followed by the use of potentials more negative than -400 mV. PMID:22610438

  15. Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

    International Nuclear Information System (INIS)

    When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal-hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal-hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.

  16. Assessment of the implications of conversion of university research and training reactors to low enrichment uranium fuel

    International Nuclear Information System (INIS)

    The tasks associated with conversion of a research reactor from HEU to LEU fuel are: initial program planning; safety analysis and license amendment; core physics calculations; operating thermal-hydraulics analysis; plant engineering modifications; LEU fuel specifications, procurement of fuel, and calculational confirmation of design; training of staff personnel; HEU core physics measurements and fuel disposal; and experimental verification of reactor behavior with LEU fuel. LEU fuel conversion of the 25 NRC licensed, university-owned reactors considered in this study is based upon the reactor fuel cycle, the type of license modification, and fuel meat technology. Reactors that operate on routine refueling cycles could periodically replace depleted HEU elements with fresh LEU elements. Ultimate full core conversion would depend on the average element residence time in the core. Reactors with lifetime cores would convert by full core replacement as a one-time event. For some reactors, LEU conversion depends upon high density uranium fuel meat technology development. The majority should be able to convert using a direct substitution of current fuel meat technology though some fuel plate or rod internal modifications may be necessary for 16 of the reactors

  17. RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A.

    1983-01-01

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

  18. Results of the work on development of research reactor fuel element based on high density fuel with decreased enrichment in uranium-235

    International Nuclear Information System (INIS)

    The work is developed to the branch program on decrease of fuel enrichment in the Russian research reactors. The analysis of results of foreign works and own studies on creating dispersion fuel compositions on the basis of high density uranium compounds is accomplished. The uranium alloys U3Si, U3Si alloys, U-Zr-Nb, U6Fe and U6Fe alloys are chosen for application in further developments. Characteristics of these alloys are given from the viewpoint of their behaviour under irradiation

  19. Effects of inoculation sources on the enrichment and performance of anode bacterial consortia in sensor typed microbial fuel cells

    Directory of Open Access Journals (Sweden)

    Phuong Tran

    2016-01-01

    Full Text Available Microbial fuel cells are a recently emerging technology that promises a number of applications in energy recovery, environmental treatment and monitoring. In this study, we investigated the effect of inoculating sources on the enrichment of electrochemically active bacterial consortia in sensor-typed microbial fuel cells (MFCs. Several MFCs were constructed, operated with modified artificial wastewater and inoculated with different microbial sources from natural soil, natural mud, activated sludge, wastewater and a mixture of those sources. After enrichment, the MFCs inoculated with the natural soil source generated higher and more stable currents (0.53±0.03 mA, in comparisons with the MFCs inoculated with the other sources. The results from denaturing gradient gel electrophoresis (DGGE showed that there were significant changes in bacterial composition from the original inocula to the enriched consortia. Even more interestingly, Pseudomonas sp. was found dominant in the natural soil source and also in the corresponding enriched consortium. The interactions between Pseudomonas sp. and other species in such a community are probably the key for the effective and stable performance of the MFCs.

  20. Dimensional stability of low enriched uranium silicide plate-type fuel for research reactors at transient conditions

    International Nuclear Information System (INIS)

    This paper describes the result of transient experiments using low enriched uranium silicide plate-type fuel for research reactors. The pulse irradiation was carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute. The results obtained were: (1) At fuel plate temperature of below 400degC, a good dimensional stability of the tested fuel was kept. No fuel failure occurred. (2) At a plate temperature of about 540degC, a local crack was initiated on the Al-3% Mg alloy cladding. Once the cladding temperature exceeded the melting point of 640degC, the fuel plate was degraded much by increased bowing and cracking of the denuded fuel meat occurred after relocation of molten Al cladding. Despite of these degradation, neither fragmentation of the fuel plate nor mechanical energy generation occurred up to the cladding temperature of 971degC. (3) At the temperatures of around 925degC, the reaction of silicide particles with molten Al in the matrix and that of cladding occurred, forming Al riched U (Al, Si) compounds and Si riched (U, Si) compounds at the outermost surface of the silicide particles. (author)

  1. Materials safeguards and accountability in the low enriched uranium conversion-fabrication sector of the fuel cycle

    International Nuclear Information System (INIS)

    Today materials accounting in the low enriched conversion-fabrication sector of the LWR fuel cycle is of increased importance. Low enriched uranium is rapidly becoming a precious metal with current dollar values in the range of one dollar per gram comparing with gold and platinum at 7-8 dollars per gram. In fact, people argue that its dollar value exceeds its safeguards value. Along with this increased financial incentive for better material control, the nuclear industry is faced with the impending implementation of international safeguards and increased public attention over its ability to control nuclear materials. Although no quantity of low enriched uranium (LEU) constitutes a practical nuclear explosive, its control is important to international safeguards because of plutonium production or further enrichment to an explosive grade material. The purpose of the paper is to examine and discuss some factors in the area of materials safeguards and accountability as they apply to the low enriched uranium conversion-fabrication sector. The paper treats four main topics: basis for materials accounting; our assessment of the proposed new IAEA requirements; adequacy of current practices; and timing and direction of future modifications

  2. Alternative Fuel Vehicle Adoption Increases Fleet Gasoline Consumption and Greenhouse Gas Emissions under United States Corporate Average Fuel Economy Policy and Greenhouse Gas Emissions Standards.

    Science.gov (United States)

    Jenn, Alan; Azevedo, Inês M L; Michalek, Jeremy J

    2016-03-01

    The United States Corporate Average Fuel Economy (CAFE) standards and Greenhouse Gas (GHG) Emission standards are designed to reduce petroleum consumption and GHG emissions from light-duty passenger vehicles. They do so by requiring automakers to meet aggregate criteria for fleet fuel efficiency and carbon dioxide (CO2) emission rates. Several incentives for manufacturers to sell alternative fuel vehicles (AFVs) have been introduced in recent updates of CAFE/GHG policy for vehicles sold from 2012 through 2025 to help encourage a fleet technology transition. These incentives allow automakers that sell AFVs to meet less-stringent fleet efficiency targets, resulting in increased fleet-wide gasoline consumption and emissions. We derive a closed-form expression to quantify these effects. We find that each time an AFV is sold in place of a conventional vehicle, fleet emissions increase by 0 to 60 t of CO2 and gasoline consumption increases by 0 to 7000 gallons (26,000 L), depending on the AFV and year of sale. Using projections for vehicles sold from 2012 to 2025 from the Energy Information Administration, we estimate that the CAFE/GHG AFV incentives lead to a cumulative increase of 30 to 70 million metric tons of CO2 and 3 to 8 billion gallons (11 to 30 billion liters) of gasoline consumed over the vehicles' lifetimes - the largest share of which is due to legacy GHG flex-fuel vehicle credits that expire in 2016. These effects may be 30-40% larger in practice than we estimate here due to optimistic laboratory vehicle efficiency tests used in policy compliance calculations.

  3. Alternative Fuel Vehicle Adoption Increases Fleet Gasoline Consumption and Greenhouse Gas Emissions under United States Corporate Average Fuel Economy Policy and Greenhouse Gas Emissions Standards.

    Science.gov (United States)

    Jenn, Alan; Azevedo, Inês M L; Michalek, Jeremy J

    2016-03-01

    The United States Corporate Average Fuel Economy (CAFE) standards and Greenhouse Gas (GHG) Emission standards are designed to reduce petroleum consumption and GHG emissions from light-duty passenger vehicles. They do so by requiring automakers to meet aggregate criteria for fleet fuel efficiency and carbon dioxide (CO2) emission rates. Several incentives for manufacturers to sell alternative fuel vehicles (AFVs) have been introduced in recent updates of CAFE/GHG policy for vehicles sold from 2012 through 2025 to help encourage a fleet technology transition. These incentives allow automakers that sell AFVs to meet less-stringent fleet efficiency targets, resulting in increased fleet-wide gasoline consumption and emissions. We derive a closed-form expression to quantify these effects. We find that each time an AFV is sold in place of a conventional vehicle, fleet emissions increase by 0 to 60 t of CO2 and gasoline consumption increases by 0 to 7000 gallons (26,000 L), depending on the AFV and year of sale. Using projections for vehicles sold from 2012 to 2025 from the Energy Information Administration, we estimate that the CAFE/GHG AFV incentives lead to a cumulative increase of 30 to 70 million metric tons of CO2 and 3 to 8 billion gallons (11 to 30 billion liters) of gasoline consumed over the vehicles' lifetimes - the largest share of which is due to legacy GHG flex-fuel vehicle credits that expire in 2016. These effects may be 30-40% larger in practice than we estimate here due to optimistic laboratory vehicle efficiency tests used in policy compliance calculations. PMID:26867100

  4. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  5. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  6. Partially Enriched U235 for Use as Fuel in Off-Site Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Larson, C. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    1953-08-10

    This is in reply to the memorandum of July 21, 1953, from R.W. Cook to S.R. Sapirie, requesting opinions on several questions involved in substituting uranium of low enrichment for the highly enriched material now used in the active lattice of the reactor in the Bulk Shielding Facility. The numbered paragraphs below correspond to the numbered paragraphs in Mr. Cook's memorandum.

  7. TRIGA fuel enrichment verification based on the measurement of short-lived fission products

    Energy Technology Data Exchange (ETDEWEB)

    Peir, J.-J.; Liu, C.-C. [Nuclear Science and Technology Development Center, National Tsing Hua University, Hsinchu, Taiwan (China); Wang, T.-K. [Department of Engineering and System Science, National Tsing Hua University, Hsinchu, Taiwan (China)

    1999-06-01

    A method is developed to verify the {sup 235}U content of TRIGA fresh fuel using gamma-ray spectrometry of the short-lived fission products {sup 97}Zr/{sup 97}Nb, {sup 132}I and {sup 140}La. The short-lived fission-product activities can be established by irradiating the fuel in a nuclear reactor. Based on the measured activities, the {sup 235}U content can be deduced by iterative calculations. The aim of this work is to establish a calibration method for estimating the burnup values of the rod-type spent fuels without the need for detailed data on fuel irradiation history.

  8. TRIGA fuel enrichment verification based on the measurement of short-lived fission products

    International Nuclear Information System (INIS)

    A method is developed to verify the 235U content of TRIGA fresh fuel using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I and 140La. The short-lived fission-product activities can be established by irradiating the fuel in a nuclear reactor. Based on the measured activities, the 235U content can be deduced by iterative calculations. The aim of this work is to establish a calibration method for estimating the burnup values of the rod-type spent fuels without the need for detailed data on fuel irradiation history

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

    International Nuclear Information System (INIS)

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

  10. Nitrogen removal in a single-chamber microbial fuel cell with nitrifying biofilm enriched at the air cathode

    KAUST Repository

    Yan, Hengjing

    2012-05-01

    Nitrogen removal is needed in microbial fuel cells (MFCs) for the treatment of most waste streams. Current designs couple biological denitrification with side-stream or combined nitrification sustained by upstream or direct aeration, which negates some of the energy-saving benefits of MFC technology. To achieve simultaneous nitrification and denitrification, without extra energy input for aeration, the air cathode of a single-chamber MFC was pre-enriched with a nitrifying biofilm. Diethylamine-functionalized polymer (DEA) was used as the Pt catalyst binder on the cathode to improve the differential nitrifying biofilm establishment. With pre-enriched nitrifying biofilm, MFCs with the DEA binder had an ammonia removal efficiency of up to 96.8% and a maximum power density of 900 ± 25 mW/m 2, compared to 90.7% and 945 ± 42 mW/m 2 with a Nafion binder. A control with Nafion that lacked nitrifier pre-enrichment removed less ammonia and had lower power production (54.5% initially, 750 mW/m 2). The nitrifying biofilm MFCs had lower Coulombic efficiencies (up to 27%) than the control reactor (up to 36%). The maximum total nitrogen removal efficiency reached 93.9% for MFCs with the DEA binder. The DEA binder accelerated nitrifier biofilm enrichment on the cathode, and enhanced system stability. These results demonstrated that with proper cathode pre-enrichment it is possible to simultaneously remove organics and ammonia in a single-chamber MFC without supplemental aeration. © 2012 Elsevier Ltd.

  11. Calculated activities of some isotopes in the RA reactor highly enriched fuel significant for possible environmental contamination - Operational report

    International Nuclear Information System (INIS)

    This report contains calculation basis and obtained results of activities for three groups of isotopes in the RA reactor 80% enriched fuel element. The following isotopes are included: 1) 85mKr, 87Kr, 88Kr, 131J, 132J, 133J, 134J, 135J, 133Xe, 138Xe i 138Cs, 2) 89Sr, 90Sr, 91Sr, 92Sr, 95Zr, 97Zr, 103Ru, 105Ru, 106Ru, 129mTe, 134Cs, 137Cs, 140Ba, 144Ce, kao i 3) 238Pu, 239Pu i 240Pu. It was estimated that the fuel is exposed to mean neutron flux. The periodicity of reactor operation is taken into account. Calculation results are given dependent on the time of exposure. These results are to be used as source data for Ra reactor safety analyses

  12. Determination of U235 enrichment from nuclear fuel by neutronic activation

    International Nuclear Information System (INIS)

    The enrichment of 235U in UO2 pellets samples through the instrumental neutron activation analysis method (I.N.A.A.) was determined. By high resolution gamma-ray spectrometry (H.R.G.S.), from analysis of isotopic ratios between fission products peaks from 235U and 239Np different energies peaks from 238U, the enrichment was achieved. The 'Boatstrap' statistics technique for the analytical results, which is based in shaping results of an unknown distribution to the Gaussian distribution by B replications in interested statistics such as: the mean and its standard error, was introduced. (M.J.C.)

  13. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  14. Selective enrichment of electrogenic bacteria for fuel cell application: Enumerating microbial dynamics using MiSeq platform.

    Science.gov (United States)

    Vamshi Krishna, K; Venkata Mohan, S

    2016-08-01

    This study is intended to examine the effect of pretreatment on selective enrichment of electrogenic bacteria from mixed culture. It has been observed that the iodopropane and heat-shock pretreatments suppress the growth of non-exoelectrons, while selecting only a limited number of strains belonging to genera Xanthomonas, Pseudomonas and Prevotella while untreated control inoculum showed more diverse community comprising of both exoelectrogens and non-exoelectrogens. High power output was observed in iodopropane (180mW/m(2)) pretreated microbial fuel cell (MFC) compared to heat-shock pretreated MFC (128mW/m(2)) and untreated control (92mW/m(2)). Coulombic efficiency of iodopropane and heat-shock pretreated MFC was higher compared to untreated control MFC, while drop in pH and volatile fatty acids (VFA) production was less in iodopropane pretreated MFC signifying the shifts in bacterial community structure toward electrogenesis instead of fermentation. These results signify the role of iodopropane and heat pretreatments on enrichment of electrogenic bacteria for fuel cell application. PMID:27061058

  15. Influence of engine speed and the course of the fuel injection characteristics on forming the average combustion temperature in the cylinder of turbo diesel engine

    Directory of Open Access Journals (Sweden)

    Piotr GUSTOF

    2007-01-01

    Full Text Available Average combustion temperatures inside a turbo diesel engine for the same load and the same total doze of fuel for two rotational speeds: 2004 [rpm] and 4250 [rpm] are presented in this paper. The aim of this work is also the evaluation of the influence of the temporary course of the fuel injection characteristics on forming temperature in theengine cylinder space for these temperatures. The calculations were carried out by means of two zone combustion model.

  16. The Melt-Dilute Treatment of Al-Base Highly Enriched DOE Spent Nuclear Fuels: Principles and Practices

    International Nuclear Information System (INIS)

    The melt-dilute treatment technology program is focused on the development and implementation of a treatment technology for diluting highly enriched (>20 percent 235U) aluminum spent nuclear fuel to low enriched levels (<20 percent 235U) and qualifying the LEU Al-SNF form for geologic repository storage. In order to reduce the enrichment of these assemblies prior to ultimate geologic repository disposal, the melt-dilute technology proposes to melt these SNF assemblies and then dilute with additions of depleted uranium. The benefits accrued from this treatment process include the potential for significant volume reduction, reduced criticality potential, and the potential for enhanced SNF form characteristics. The emphasis within the development program to date has been on determining the process metallurgy and off-gas system design for the treatment of all types of Al SNF (UAlx, Al-U3O8, and Al-U3Si2). In determining the process metallurgy a wide range of alloys, representative of those expected in the Al-SNF form, have been fabricated and their product characteristics, namely microstructure, homogeneity, phase composition, and ''ternary'' constituent effects have been analyzed. As a result of the presence of species within the melt which will possess significant vapor pressures in the desired operating temperature range an off-gas system is necessary. Of the volitile species the one of greatest concern is 137Cs

  17. The Melt-Dilute Treatment of Al-Base Highly Enriched DOE Spent Nuclear Fuels: Principles and Practices

    Energy Technology Data Exchange (ETDEWEB)

    Adams, T.M.

    1998-11-25

    The melt-dilute treatment technology program is focused on the development and implementation of a treatment technology for diluting highly enriched (>20 percent 235U) aluminum spent nuclear fuel to low enriched levels (<20 percent 235U) and qualifying the LEU Al-SNF form for geologic repository storage. In order to reduce the enrichment of these assemblies prior to ultimate geologic repository disposal, the melt-dilute technology proposes to melt these SNF assemblies and then dilute with additions of depleted uranium. The benefits accrued from this treatment process include the potential for significant volume reduction, reduced criticality potential, and the potential for enhanced SNF form characteristics. The emphasis within the development program to date has been on determining the process metallurgy and off-gas system design for the treatment of all types of Al SNF (UAlx, Al-U3O8, and Al-U3Si2). In determining the process metallurgy a wide range of alloys, representative of those expected in the Al-SNF form, have been fabricated and their product characteristics, namely microstructure, homogeneity, phase composition, and "ternary" constituent effects have been analyzed. As a result of the presence of species within the melt which will possess significant vapor pressures in the desired operating temperature range an off-gas system is necessary. Of the volitile species the one of greatest concern is 137Cs.

  18. Development of low enrichment technologies for high density fuels and for isotope production targets

    International Nuclear Information System (INIS)

    Since more than twenty years ago, CNEA has carried out RERTR activities. Main goals are to convert the RA 6 reactor core from HEU to LEU, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, and to optimize techniques to recover U from silicide scrap samples. The future plans include: 1) Completion the RA 6 reactor conversion to LEU; 2) Qualification by irradiation of the promising solutions found for the high density fuels; 3) Irradiation of mini plates and full scale fuel assemblies at the RA 3 reactor and at higher flux and temperature reactors; 4) Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  19. Characterization of two diesel fuel degrading microbial consortia enriched from a non acclimated, complex source of microorganisms

    Directory of Open Access Journals (Sweden)

    Varese Giovanna C

    2010-02-01

    Full Text Available Abstract Background The bioremediation of soils impacted by diesel fuels is very often limited by the lack of indigenous microflora with the required broad substrate specificity. In such cases, the soil inoculation with cultures with the desired catabolic capabilities (bioaugmentation is an essential option. The use of consortia of microorganisms obtained from rich sources of microbes (e.g., sludges, composts, manure via enrichment (i.e., serial growth transfers on the polluting hydrocarbons would provide bioremediation enhancements more robust and reproducible than those achieved with specialized pure cultures or tailored combinations (co-cultures of them, together with none or minor risks of soil loading with unrelated or pathogenic allocthonous microorganisms. Results In this work, two microbial consortia, i.e., ENZ-G1 and ENZ-G2, were enriched from ENZYVEBA (a complex commercial source of microorganisms on Diesel (G1 and HiQ Diesel (G2, respectively, and characterized in terms of microbial composition and hydrocarbon biodegradation capability and specificity. ENZ-G1 and ENZ-G2 exhibited a comparable and remarkable biodegradation capability and specificity towards n-C10 to n-C24 linear paraffins by removing about 90% of 1 g l-1 of diesel fuel applied after 10 days of aerobic shaken flask batch culture incubation at 30°C. Cultivation dependent and independent approaches evidenced that both consortia consist of bacteria belonging to the genera Chryseobacterium, Acinetobacter, Psudomonas, Stenotrophomonas, Alcaligenes and Gordonia along with the fungus Trametes gibbosa. However, only the fungus was found to grow and remarkably biodegrade G1 and G2 hydrocarbons under the same conditions. The biodegradation activity and specificity and the microbial composition of ENZ-G1 and ENZ-G2 did not significantly change after cryopreservation and storage at -20°C for several months. Conclusions ENZ-G1 and ENZ-G2 are very similar highly enriched consortia

  20. A Neutronic Analysis of TRU Recycling in PWRs Loaded with MOX-UE Fuel (MOX with U-235 Enriched U Support)

    Energy Technology Data Exchange (ETDEWEB)

    G. Youinou; S. Bays

    2009-05-01

    This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study).

  1. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  2. Uranium enrichment

    International Nuclear Information System (INIS)

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  3. Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

    International Nuclear Information System (INIS)

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub ∞/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub ∞/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub ∞/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs

  4. Direct Measurement of Initial Enrichment, Burn-up and Cooling Time of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    Energy Technology Data Exchange (ETDEWEB)

    Henzl, Vladimir [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

    2012-07-13

    An outline of this presentation of what a Differential Die-Away (DDA) instrument can do are: (1) Principle of operation of DDA instrument; (2) Determination of initial enrichment (IE) ({sigma} < 5%); (3) Determination of burn up (BU) ({sigma} {approx} 6%); (4) Determination of cooling time (CT) ({sigma} {approx} 20-50%); and (5) DDA instrument as a standalone device. DDA response (fresh fuel vs. spent fuel) is: (1) Fresh fuel => DDA response increases (die-away time is longer) with increasing fissile content; and (2) Spent fuel => DDA response decreases (die-away time is shorter) with higher burn-up (i.e. more neutron absorbers present).

  5. Detailed description of an SSAC at the facility level for a low-enriched uranium conversion and fuel fabrication facility

    International Nuclear Information System (INIS)

    Some States have expressed a need for more detailed guidance with regard to the technical elements in the design and operation of SSACs for both the national and the international objectives. To meet this need the present document has been prepared, describing the technical elements of an SSAC in considerable detail. The purpose of this document is therefore, to provide a detailed description of a system for the accounting for and control of nuclear material in a model low enriched uranium conversion and fuel fabrication facility which can be used by a facility operator to establish his own system in a way which will provide the necessary information for compliance with a national system for nuclear material accounting and control and for the IAEA to carry out its safeguards responsibilities

  6. Average Droplet Diameter Measurement and Results for Fuel Aerosol Injected by Certain Types of the Turbojet Burners

    Institute of Scientific and Technical Information of China (English)

    TadeuszOpara

    1997-01-01

    Measurement of the diameter of the fuel aerosol droplet is very important in the design of new type burners and in diagnostic process,Diffraction method is one of the most useful measuring procedures in this case.An investigation setup is presented enabling the determination of the substituting drop diameter in fuel aerosol stream created by aeroengine injectors the results obtained for K 108-767,K 108-012,37.03.9595,16.83.0310 types are presented.

  7. Monte Carlo calculational design of an NDA instrument for the assay of waste products from high enriched uranium spent fuels

    International Nuclear Information System (INIS)

    The Monte Carlo design of the waste assay region of a dual assay system, to be installed at the Fluorinal and Storage Facility, is described. The instrument will be used by the facility operator to assay high-enriched spent fuel packages and waste solids produced from dissolution of the fuels. The fissile content discharged in the waste is expected to vary between 0 and 400 g of 235U. Material accountability measurements of the waste must be obtained in the presence of large neutron (0.5 x 106 n/s) and gamma (50,000 R/hr) backgrounds. The assay system employs fast-neutron irradiation of the sample, using a 5 mg 252Cf source, followed by delayed neutron counting after the source is transferred to storage. Calculations indicate a +-4-g (2 sigma) assay for a waste canister containing 300 g of 235U is achievable with an end-of-life (1 mg) 252Cf source and a background rate of 0.5 x 106 n/s

  8. Calculation of average molecular parameters, functional groups, and a surrogate molecule for heavy fuel oils using 1H and 13C NMR spectroscopy

    KAUST Repository

    Abdul Jameel, Abdul Gani

    2016-04-22

    Heavy fuel oil (HFO) is primarily used as fuel in marine engines and in boilers to generate electricity. Nuclear Magnetic Resonance (NMR) is a powerful analytical tool for structure elucidation and in this study, 1H NMR and 13C NMR spectroscopy were used for the structural characterization of 2 HFO samples. The NMR data was combined with elemental analysis and average molecular weight to quantify average molecular parameters (AMPs), such as the number of paraffinic carbons, naphthenic carbons, aromatic hydrogens, olefinic hydrogens, etc. in the HFO samples. Recent formulae published in the literature were used for calculating various derived AMPs like aromaticity factor 〖(f〗_a), C/H ratio, average paraffinic chain length (¯n), naphthenic ring number 〖(R〗_N), aromatic ring number〖 (R〗_A), total ring number〖 (R〗_T), aromatic condensation index (φ) and aromatic condensation degree (Ω). These derived AMPs help in understanding the overall structure of the fuel. A total of 19 functional groups were defined to represent the HFO samples, and their respective concentrations were calculated by formulating balance equations that equate the concentration of the functional groups with the concentration of the AMPs. Heteroatoms like sulfur, nitrogen, and oxygen were also included in the functional groups. Surrogate molecules were finally constructed to represent the average structure of the molecules present in the HFO samples. This surrogate molecule can be used for property estimation of the HFO samples and also serve as a surrogate to represent the molecular structure for use in kinetic studies.

  9. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    International Nuclear Information System (INIS)

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s−1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  10. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  11. HTGR Generic Technology Program. Materials technology reactor operating experience medium-enriched-uranium fuel development. Quarterly progress report for the period ending April 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Kaae, J. L.; Lai, G. Y.; Thompson, L. D.; Sheehan, J. E.; Rosenwasser, S. N.; Johnson, W. R.; Li, C. C.; Pieren, W. R.; Smith, A. B.; Holko, K. H.; Baenteli, G. J.; Cheung, K. C.; Orr, J. D.; Potter, R. C.; Baxter, A.; Bell, W.; Lane, R.; Wunderlich, R. G.; Neylan, A. J.

    1978-05-01

    The work reported includes the development of the materials properties data base for noncore components, plant surveillance and testing performed at Fort St. Vrain, and work to demonstrate the feasibility of using medium-enriched fuel in Fort St. Vrain. Studies and analyses plus experimental procedures and results are discussed and data are presented.

  12. HTGR Generic Technology Program: materials technology reactor; operating experience; medium-enriched-uranium fuel development. Quarterly progress report for the period ending July 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    The work reported includes the development of the materials properties data base for noncore components, plant surveillance and testing performed at Fort St. Vrain, and work to demonstrate the feasibility of using medium-enriched fuel in Fort St. Vrain. Studies and analyses plus experimental procedures and results are discussed and data are presented.

  13. 77 FR 62623 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Science.gov (United States)

    2012-10-15

    ...EPA and NHTSA, on behalf of the Department of Transportation, are issuing final rules to further reduce greenhouse gas emissions and improve fuel economy for light-duty vehicles for model years 2017 and beyond. On May 21, 2010, President Obama issued a Presidential Memorandum requesting that NHTSA and EPA develop through notice and comment rulemaking a coordinated National Program to improve......

  14. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Science.gov (United States)

    2010-07-01

    ... Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.60... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false How do we calculate...

  15. 75 FR 25323 - Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards...

    Science.gov (United States)

    2010-05-07

    ...-by-the-President-on-national-fuel-efficiency-standards/ . \\4\\ 74 FR 24007 (May 22, 2009). \\5... reconsideration in light of the Court's decision.\\17\\ \\14\\ 549 U.S. 497 (2007). \\15\\ 68 FR 52922 (Sept. 8, 2003... FR 44354 at 44397. There is a comprehensive discussion of the litigation's history, the Supreme...

  16. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  17. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  18. Safety evaluation report related to the evaluation of low-enriched uranium silicide-aluminum dispersion fuel for use in non-power reactors

    Energy Technology Data Exchange (ETDEWEB)

    1988-07-01

    Low-enriched uranium silicide-aluminum dispersion plate-type fuels have been extensively researched and developed under the international program, Reduced Enrichment in Research and Test Reactors. The international effort was led by Argonne National Laboratory (ANL) in the United States. This evaluation is based primarily on reports issued by ANL that discuss and summarize the developmental tests and experiments, including postirradiation examinations, of both miniature and full-sized plates of prototypical fuel compositions. This evaluation concludes that plate-type fuels suitable and acceptable for use in research and test reactors can be fabricated with U/sub 3/Si/sub 2/-Al dispersion compacts with uranium densities up to 4.8 g/cm/sup 3/. 4 refs., 1 fig.

  19. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  20. Oscillator measurements of the reactivity changes resulting from the irradiation of low enrichment particulate fuel in the Dragon reactor

    International Nuclear Information System (INIS)

    This Report describes a series of experiments carried out as a joint UKAEA/CEA/DRAGON project to determine the reactivity changes of low-enrichment particulate fuel samples following their irradiation in the DRAGON reactor to various levels up to approximately 60,000 MWD/Te. The samples are described, together with the method of measurement of reactivity in the Winfrith reactor HECTOR, which was an extension of the well-known Oscillator Technique to yield simultaneously overall reactivity changes and changes in macroscopic absorption cross-sections. Measurements were carried out at room temperature in two reactor spectra; a thermal spectrum and one typical of an HTR type reactor. The resultant reactivity changes are presented together with the relevant sample burn-ups as determined by #betta#-scanning methods and, in some cases, by rigorous chemical analysis. The results of supporting measurements are also reported, carried out to characterise the neutron spectra in which the oscillator measurements were made and to determine the neutron flux distributions in the HECTOR reactor. (author)

  1. Proceedings of the international meeting on development, fabrication, and application of Reduced Enrichment fuels for Research and Test Reactors (RERTR). Base technology

    International Nuclear Information System (INIS)

    The international effort to develop new fuel materials and designs which will make it feasible to fuel research and test reactors throughout the world with low-enrichment uranium, instead of high-enrichment uranium, has made significant progress during the past year. This progress has taken place at research centers located in many different countries, and is of crucial interest to reactor operators and licensors whose geographical distribution is even more varied. It is appropriate, therefore, that international meetings be held periodically to foster direct communication among the specialists in this area. To achieve this purpose, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the third of a series which begun in 1978. The papers presented at this meeting were divided into sessions according to relevant subject: status of RERTR program and safety issues; development of new fuel types; testing of new fuel elements; specific reactor applications. These proceedings were edited by various members of the RERTR Program

  2. Partitioning of metal species during an enriched fuel combustion experiment. speciation in the gaseous and particulate phases.

    Science.gov (United States)

    Pavageau, Marie-Pierre; Morin, Anne; Seby, Fabienne; Guimon, Claude; Krupp, Eva; Pécheyran, Christophe; Poulleau, Jean; Donard, Olivier F X

    2004-04-01

    Combustion processes are the most important source of metal in the atmosphere and need to be better understood to improve flue gas treatment and health impact studies. This combustion experiment was designed to study metal partitioning and metal speciation in the gaseous and particulate phases. A light fuel oil was enriched with 15 organometallic compounds of the following elements: Pb, Hg, As, Cu, Zn, Cd, Se, Sn, Mn, V, Tl, Ni, Co, Cr, and Sb. The resulting mixture was burnt in a pilot-scale fuel combustion boiler under controlled conditions. After filtration of the particles, the gaseous species were sampled in the stack through a heated sampling tube simultaneously by standardized washing bottles-based sampling techniques and cryogenically. The cryogenic samples were collected at -80 degrees C for further speciation analysis by LT/GC-ICPMS. Three species of selenium and two of mercury were evidenced as volatile species in the flue gas. Thermodynamic predictions and experiments suggest the following volatile metal species to be present in the flue gas: H2Se, CSSe, CSe2, SeCl2, Hg(0), and HgCl2. Quantification of volatile metal species in comparison between cryogenic techniques and the washing bottles-based sampling method is also discussed. Concerning metal partitioning, the results indicated that under these conditions, at least 60% (by weight) of the elements Pb, Sn, Cu, Co, Tl, Mn, V, Cr, Ni, Zn, Cd, and Sb mixed to the fuel were found in the particulate matter. For As and Se, 37 and 17%, respectively, were detected in the particles, and no particulate mercury was found. Direct metal speciation in particles was performed by XPS allowing the determination of the oxidation state of the following elements: Sb(V), Tl(III), Mn(IV), Cd(II), Zn(II), Cr(III), Ni(II), Co(II), V(V), and Cu(II). Water soluble species of inorganic Cr, As, and Se in particulate matter were determined by HPLC/ICP-MS and identified in the oxidation state Cr(III), As(V), and Se(IV). PMID

  3. Qualification of high-density low-enriched U{sub 3}Si{sub 2} fuel for material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chaussy, J.M. [CEA/DRN/DRE, CEA Grenoble, Grenoble (France); Beylot, J.P. [CEA/DRN/DRE, CEA Saclay, Gif sur Yvette (France)

    1999-07-01

    The CEA, in collaboration with the CERCA, decided at the beginning of the 1990s to set up a programme for qualification of a high-density fuel based on a U{sub 3}Si{sub 2} alloy 19.75% enriched with {sup 235}U. The objective aimed for is to achieve a {sup 235}U density increase of 30%. This programme was brought up in discussion at the RRFM97 meeting and also in the course of RERTR meetings. This presentation sets out the channels explored and the results obtained for the qualification of experimental fuel plates irradiated. (author)

  4. 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)

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

    2006-02-01

    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.

  5. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ... Reactor (BWR) fuel with high initial enrichment (up to 4.8 weight percent uranium-235 planer average...) The ability to store and transport BWR fuel with high initial enrichment (up to 4.8 weight percent... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR...

  6. Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

    Science.gov (United States)

    Graven, H. D.; Gruber, N.

    2011-12-01

    The 14C-free fossil carbon added to atmospheric CO2 by combustion dilutes the atmospheric 14C/C ratio (Δ14C), potentially providing a means to verify fossil CO2 emissions calculated using economic inventories. However, sources of 14C from nuclear power generation and spent fuel reprocessing can counteract this dilution and may bias 14C/C-based estimates of fossil fuel-derived CO2 if these nuclear influences are not correctly accounted for. Previous studies have examined nuclear influences on local scales, but the potential for continental-scale influences on Δ14C has not yet been explored. We estimate annual 14C emissions from each nuclear site in the world and conduct an Eulerian transport modeling study to investigate the continental-scale, steady-state gradients of Δ14C caused by nuclear activities and fossil fuel combustion. Over large regions of Europe, North America and East Asia, nuclear enrichment may offset at least 20% of the fossil fuel dilution in Δ14C, corresponding to potential biases of more than -0.25 ppm in the CO2 attributed to fossil fuel emissions, larger than the bias from plant and soil respiration in some areas. Model grid cells including high 14C-release reactors or fuel reprocessing sites showed much larger nuclear enrichment, despite the coarse model resolution of 1.8°×1.8°. The recent growth of nuclear 14C emissions increased the potential nuclear bias over 1985-2005, suggesting that changing nuclear activities may complicate the use of Δ14C observations to identify trends in fossil fuel emissions. The magnitude of the potential nuclear bias is largely independent of the choice of reference station in the context of continental-scale Eulerian transport and inversion studies, but could potentially be reduced by an appropriate choice of reference station in the context of local-scale assessments.

  7. Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

    Directory of Open Access Journals (Sweden)

    N. Gruber

    2011-12-01

    Full Text Available The 14C-free fossil carbon added to atmospheric CO2 by combustion dilutes the atmospheric 14C/C ratio (Δ14C, potentially providing a means to verify fossil CO2 emissions calculated using economic inventories. However, sources of 14C from nuclear power generation and spent fuel reprocessing can counteract this dilution and may bias 14C/C-based estimates of fossil fuel-derived CO2 if these nuclear influences are not correctly accounted for. Previous studies have examined nuclear influences on local scales, but the potential for continental-scale influences on Δ14C has not yet been explored. We estimate annual 14C emissions from each nuclear site in the world and conduct an Eulerian transport modeling study to investigate the continental-scale, steady-state gradients of Δ14C caused by nuclear activities and fossil fuel combustion. Over large regions of Europe, North America and East Asia, nuclear enrichment may offset at least 20% of the fossil fuel dilution in Δ14C, corresponding to potential biases of more than −0.25 ppm in the CO2 attributed to fossil fuel emissions, larger than the bias from plant and soil respiration in some areas. Model grid cells including high 14C-release reactors or fuel reprocessing sites showed much larger nuclear enrichment, despite the coarse model resolution of 1.8°×1.8°. The recent growth of nuclear 14C emissions increased the potential nuclear bias over 1985–2005, suggesting that changing nuclear activities may complicate the use of Δ14C observations to identify trends in fossil fuel emissions. The magnitude of the potential nuclear bias is largely independent of the choice of reference station in the context of continental-scale Eulerian transport and inversion studies, but could potentially be reduced by an appropriate choice of reference station in the context of local-scale assessments.

  8. State Averages

    Data.gov (United States)

    U.S. Department of Health & Human Services — A list of a variety of averages for each state or territory as well as the national average, including each quality measure, staffing, fine amount and number of...

  9. Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

    Directory of Open Access Journals (Sweden)

    N. Gruber

    2011-05-01

    Full Text Available Since aged carbon in fossil fuel contains no 14C, 14C/C ratios (Δ14C measured in atmospheric CO2 can be used to estimate CO2 added by combustion and, potentially, provide verification of fossil CO2 emissions calculated using economic inventories. Sources of 14C from nuclear power generation and spent fuel reprocessing can counteract dilution by fossil CO2. Therefore, these nuclear sources can bias observation-based estimates of fossil fuel-derived CO2 if they are not correctly accounted for or included as a source of uncertainty. We estimate annual 14C emissions from each nuclear site in the world and conduct an Eulerian transport modeling study to investigate the continental-scale, steady-state gradients of Δ14C caused by nuclear activities and fossil fuel combustion. Over Europe, North America and East Asia, nuclear enrichment may offset 0–260 % of the fossil fuel dilution in Δ14C, corresponding to potential biases of 0 to −8 ppm in the CO2 attributed to fossil fuel emissions, larger than the bias from respiration in some areas. Growth of 14C emissions increased the potential nuclear bias over 1985–2005. The magnitude of this potential bias is largely independent of the choice of reference station in the context of Eulerian transport and inversion studies, but could potentially be reduced by an appropriate choice of reference station in the context of local-scale assessments.

  10. Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

    International Nuclear Information System (INIS)

    Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

  11. Treatment and electricity harvesting from sulfate/sulfide-containing wastewaters using microbial fuel cell with enriched sulfate-reducing mixed culture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duu-Jong, E-mail: cedean@mail.ntust.edu.tw [Department of Chemical Engineering, National Taiwan University, Taipei, Taiwan (China); Department of Chemical Engineering, National Taiwan University of Science and Technology, Taipei, Taiwan (China); Lee, Chin-Yu [Department of Chemical Engineering, National Taiwan University, Taipei, Taiwan (China); Chang, Jo-Shu [Department of Chemical Engineering, National Cheng Kung University, Tainan, Taiwan (China); Center for Bioscience and Biotechnology, National Cheng Kung University, Tainan, Taiwan (China); Research Center for Energy Technology and Strategy, National Cheng Kung University, Tainan, Taiwan (China)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer We started up microbial fuel cell (MFC) using enriched sulfate-reducing mixed culture. Black-Right-Pointing-Pointer Sulfate-reducing bacteria and anode-respiring bacteria were enriched in anodic biofilms. Black-Right-Pointing-Pointer The MFC effectively remove sulfate to elementary sulfur in the presence of lactate. Black-Right-Pointing-Pointer The present device can treat sulfate laden wastewaters with electricity harvesting. - Abstract: Anaerobic treatment of sulfate-laden wastewaters can produce excess sulfide, which is corrosive to pipelines and is toxic to incorporated microorganisms. This work started up microbial fuel cell (MFC) using enriched sulfate-reducing mixed culture as anodic biofilms and applied the so yielded MFC for treating sulfate or sulfide-laden wastewaters. The sulfate-reducing bacteria in anodic biofilm effectively reduced sulfate to sulfide, which was then used by neighboring anode respiring bacteria (ARB) as electron donor for electricity production. The presence of organic carbons enhanced MFC performance since the biofilm ARB were mixotrophs that need organic carbon to grow. The present device introduces a route for treating sulfate laden wastewaters with electricity harvesting.

  12. World nuclear-fuel procurement: relationships between uranium and enrichment markets. Final report. International energies studies program

    International Nuclear Information System (INIS)

    This article explores the relationships between international uranium and enrichment markets under current contracting and equity arrangements and in comparison with actual feed requirements for existing and committed reactors. We begin with an overview of the world situation, examining current and prospective conditions. We then consider enrichment and uranium supply and demand situations of the three consumer nations outside the United States with the largest nuclear programs: France, Japan, and the Federal Republic of Germany. We conclude with an evaluation of likely directions of change in the coupled markets for uranium and enrichment services

  13. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  14. Quaternion Averaging

    Science.gov (United States)

    Markley, F. Landis; Cheng, Yang; Crassidis, John L.; Oshman, Yaakov

    2007-01-01

    Many applications require an algorithm that averages quaternions in an optimal manner. For example, when combining the quaternion outputs of multiple star trackers having this output capability, it is desirable to properly average the quaternions without recomputing the attitude from the the raw star tracker data. Other applications requiring some sort of optimal quaternion averaging include particle filtering and multiple-model adaptive estimation, where weighted quaternions are used to determine the quaternion estimate. For spacecraft attitude estimation applications, derives an optimal averaging scheme to compute the average of a set of weighted attitude matrices using the singular value decomposition method. Focusing on a 4-dimensional quaternion Gaussian distribution on the unit hypersphere, provides an approach to computing the average quaternion by minimizing a quaternion cost function that is equivalent to the attitude matrix cost function Motivated by and extending its results, this Note derives an algorithm that deterniines an optimal average quaternion from a set of scalar- or matrix-weighted quaternions. Rirthermore, a sufficient condition for the uniqueness of the average quaternion, and the equivalence of the mininiization problem, stated herein, to maximum likelihood estimation, are shown.

  15. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2014-12-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  16. Computation of concentration changes of heavy metals in the fuel assemblies with 1.6% enrichment by ORIGEN code for VVER-1000

    International Nuclear Information System (INIS)

    ORIGEN code is a widely used computer code for calculating the buildup, decay, and processing of radioactive materials. During the past few years, a sustained effort was undertaken by ORNL to update the original ORIGEN code [4] and its associated data bases. The results of this effort were updated on the reactor model, cross section, fission product yields, decay data, decay photon data and the ORIGEN computer code itself. In this paper we have obtained concentration changes of uranium and plutonium isotopes by ORIGEN code at different burn-up and then the results have been compared with VVER-1000 results in the first fuel cycle for fuel assemblies with 1.6% enrichment in the BUSHEHR Nuclear Power Plant. (author)

  17. Thermal-hydraulic analysis for core conversion to the use of low-enriched uranium fuels in the KUR

    International Nuclear Information System (INIS)

    Thermal-hydraulic analysis was made for LEU fuels in the KUR. Five fuel element geometries are studied. Their dimensions are assumed as combinations of the following parameters, keeping the outer dimensions of element unchanged: meat thickness = 0.50 and 0.61mm, clad thickness = 0.51 and 0.38mm, channel thickness = 2.81 and 2.20mm. Fuel plates are assumed to be either curved or flat and the number of fuel plates ranged from 18 to 22 according to the change in the dimensions. For each fuel element, the relation between the pressure loss and the flow rate, critical heat flux, and heat fluxes for the onset of flow instability and the onset of nucleate boiling are calculated using the computer code PLTEMP3 which has been developed for this study. The effect of fuel material (UAlsub(x)-Al, U3O8-Al and U3Si2-Al) on the peak fuel temperature is also studied. As a particular interest in the mixed core which may be constructed on the way to the use of LEU fuels, the change in the bypass flow rate due to the change in the gap between different fuel elements is investigated. The results indicate that no significant problem arises due to core conversion for the use of LEU fuels in view of core thermal-hydraulics. With the assumptions used in the analysis and assuming that the permissible minimum margin to ONB be 1.2, the total peaking factor should be lower than 3.3 as far as the primary cooling system is unchanged. (author)

  18. A sensitivity study on DUPIC fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong

    1997-01-01

    In DUPIC fuel cycle, the spent pressurized water reactor (PWR) fuel is refabricated as a DUPIC fuel by a dry process. Because the spent PWR fuel composition depends on the initial enrichment and burnup condition of PWR fuel, the composition of a DUPIC fuel is not uniquely defined. Therefore, for the purpose of reducing the effects of such a composition heterogeneity on core performance, a composition adjustment of DUPIC fuel was studies. The composition adjustment was made in two steps: mixing two spent PWR fuel assemblies of higher and lower {sup 239}Pu contents and blending in fresh uranium with the mixed spent PWR fuels. Because the fuel and core performances depend on both the absolute amount of fissile isotopes and the ratio of major fissile isotope contents, a parametric study was performed to determine the reference compositions of {sup 235}U and {sup 239}Pu. The reference enrichments of {sup 235}U and {sup 239}Pu were determined such that the DUPIC core performance is comparable to that of a natural uranium core with high spent PWR fuel utilization and low fuel cycle cost. Under this condition, it is possible to utilize 90% of spent PWR fuels as the DUPIC fuel formula. On average, the amounts of slightly enriched and depleted uranium used for blending correspond to 8.6% and 10.6%, respectively, of the mass of candidate spent PWR fuels. (author). 16 refs., 30 tabs., 9 figs.

  19. The effect of alternative fuel combustion in the cement kiln main burner on production capacity and improvement with oxygen enrichment.

    OpenAIRE

    Ariyaratne, W. K. Hiromi; Melaaen, Morten Christian; Tokheim, Lars-André

    2013-01-01

    A mathematical model based on a mass and energy balance for the combustion in a cement rotary kiln was developed. The model was used to investigate the impact of replacing about 45 % of the primary coal energy by different alternative fuels. Refuse derived fuel, waste wood, solid hazardous waste and liquid hazardous waste were used in the modeling. The results showed that in order to keep the kiln temperature unchanged, and thereby maintain the required clinker quality, the production capa...

  20. Establishing a quality assurance program for in-core fuel management of the Dalat Nuclear Research Reactor using low enriched fuel

    International Nuclear Information System (INIS)

    Quality assurance program for calculating of in-core fuel management of research reactor plays very important role in safety operation and effective utilization. The main objective of the program is to ensure the safe, reliable and optimum use of nuclear fuel and to meet the reactor utilization, which remains reactor operation within the limits imposed by the design safety considerations and the operational limits and conditions (OLCs) on the basis of safety analysis. The management of reactor core and nuclear fuel must be organized in a coherent way and comply with safety requirements. After successfully converting from HEU to LEU fuel for Dalat Research Reactor, a work to be in place is to study and implement the management of reactor core and nuclear fuel. This not only helps to ensure safety operation and efficient utilization but also contributes to build the safety culture and to be valuable experience for other nuclear projects. In addition, the application of the quality assurance program for in-core fuel management will contribute to avoid subjective mistakes, to clearly define responsibilities and to ensure legacy of expertise, which is also an urgent requirement. The selected computer code systems, data libraries and computation models must be fully met the requirements for analyzing status and characteristics of reactor core as well as the requirements for selecting, verifying and evaluating according to the regulations of the IAEA. (author)

  1. In-plane and through-plane local and average Nusselt numbers in fibrous porous materials with different fiber layer temperatures: Gas diffusion layers for fuel cells

    Science.gov (United States)

    Sadeghifar, Hamidreza

    2016-09-01

    Convective heat transfer inside fibrous gas diffusion layers (GDLs) noticeably impacts the heat and water management of air-cooled polymer electrolyte membrane fuel cells (PEMFCs). Cutting-edge experiments have recently proved that convective heat transfer inside fibrous GDLs increases their thermal resistances considerably. However, heat transfer coefficients are difficult to measure experimentally or compute numerically for the millions of the tiny pores inside microstructural GDLs. The present study provides robust analytic models for predicting the heat transfer coefficient for both through-plane and in-plane flows inside fibrous media such as GDLs. The model is based on the unit cell approach and the integral method. Closed-form formulas are developed for local and average heat transfer coefficients. The model considers the temperature variations of the fiber layers along the medium thickness while assuming the same temperature for all the fibers in each layer. The model is well verified by COMSOL numerical data for a few pores inside a GDL. The simple, closed-form easy-to-use formulas developed in this study can be readily employed for predicting Nusselt number inside multilayer fibrous porous materials.

  2. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  3. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant

    International Nuclear Information System (INIS)

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. Then a statistical inference can be made from verification results for items verified during SNRIs to the entire populations, i.e. the entire strata, even if inspectors were not present when many items were received or produced. A six-month field test of the feasibility of such SNRIs took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division during 1993. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ''mailbox''. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. They arrived unannounced at the plant, in most cases immediately after travel from Canada, where the IAEA maintains a regional office. Items from both strata were verified during the SNRIs by meant of nondestructive assay equipment

  4. Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment

    International Nuclear Information System (INIS)

    The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author)

  5. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  6. Investigative studies on the effects of cadmium rabbits on high enriched uranium-fueled and low enriched uranium-fueled cores of Ghana Research Reactor-1 using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Boffie, J., E-mail: jboffie@yahoo.com [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Akaho, E.H.K. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Nyarko, B.J.B.; Odoi, H.C.; Tuffour-Achampong, K.; Abrefah, R.G. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana)

    2013-12-15

    Highlights: • The operating parameters for both the HEU core and proposed LEU core were similar. • The length of the Cd in the capsules must be increased for its use in the LEU core. • Cd rabbits can emergently be used to shut down MNSRs. - Abstract: Miniature Neutron Source Reactors (MNSRs) are noted to be among highly safe research reactors. However, because of its use of one control rod for reactivity control and shutdown purposes, alternative methods of shutting it down are important. The Ghana MNSR uses four cadmium rabbits of approximate dimensions 6.5 cm × 5.0 cm × 0.1 cm and mass of 9.48 g each to emergently shut down the reactor. The Monte Carlo N-Particle code; version 5, (MCNP5) was used to design the high enriched uranium (HEU) and low enriched uranium (LEU) cores of the MNSR with four cadmium rabbits inserted in four inner irradiation sites of each core. The operating parameters and shutdown parameters for both cores with the central control rod (CCR) either fully withdrawn or fully inserted had similar results with the HEU core having slightly better results in terms of safety. However, the results show that the four inserted cadmium rabbits make the HEU core subcritical whiles in the LEU core, it still remains critical (k{sub eff} = 1.00005 ± 0.00007). The length of the cadmium material in each cadmium rabbit must therefore be increased by at least 0.5 cm in order to attain subcriticality (k{sub eff} = 0.99989 ± 0.00006) and shutdown margin of 0.11 mk when inserted in the LEU core.

  7. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL; Curtis, Franklin G [ORNL; Arimilli, Rao V [ORNL; Ekici, Kivanc [ORNL; Freels, James D [ORNL

    2015-12-01

    ABSTRACT The findings presented in this report are results of a five year effort lead by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor

  8. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

    ABSTRACT The findings presented in this report are results of a five year effort lead by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor

  9. First shipment of TRIGA 14MW research reactor highly enriched uranium spent fuel to the United States of America

    International Nuclear Information System (INIS)

    The TRIGA 14MW Research Reactor has a unique design of core and fuel, with an exceptionally long life. This means long time in-core utilization, leading to a high burnup. The peculiar characteristics of the fuel and reactor facility design made the first shipment dissimilar from the other TRIGA reactors or aluminium plate type shipments. The paper presents the legal framework, regulatory activity, licensing, agreements, contracts, training prior to shipment. The shipment was considered a large coordinated project requiring preparatory activities, resources, national and international cooperation. The overall project time schedule is presented, as well as the diagram of the activities with intervening groups, organization and logistics, the unforeseen events being also mentioned. (author)

  10. Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

    International Nuclear Information System (INIS)

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U3Si and U3Si2) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U3Si- und U3Si-Al up to U-densities of 6.0 g U/cm3. The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.)

  11. Direct Measurement of Initial Enrichment and Burn-up of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    Energy Technology Data Exchange (ETDEWEB)

    Henzl, Vladimir [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

    2012-07-16

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to utilize non-destructive assay (NDA) techniques to determine the elemental plutonium (Pu) content in a commercial-grade nuclear spent fuel assembly (SFA). In the third year of the NGSI Spent Fuel NDA project, the research focus is on the integration of a few NDA techniques. One of the reoccurring challenges to the accurate determination of Pu content has been the explicit dependence of the measured signal on the presence of neutron absorbers which build up in the assembly in accordance with its operating and irradiation history. The history of any SFA is often summarized by the parameters of burn-up (BU), initial enrichment (IE) and cooling time (CT). While such parameters can typically be provided by the operator, the ability to directly measure and verify them would significantly enhance the autonomy of the IAEA inspectorate. Within this paper, we demonstrate that an instrument based on a Differential Die-Away technique is in principle capable of direct measurement of IE and, should the CT be known, also the BU.

  12. 鼓泡流化床垃圾衍生燃料富氧气化%Enriched-air gasification of refuse derived fuel in bubbling fluidized bed

    Institute of Scientific and Technical Information of China (English)

    牛淼淼; 黄亚继; 金保昇; 王妍艳; 董新新

    2014-01-01

    Enriched air gasification of two different refuse derived fuels (RDF) was performed in a bubbling fluidized bed reactor. Thermo-gravimetric analysis of the two RDFs was performed and the effect of temperature, equivalence ratio (ER) and oxygen percentage of enriched air was investigated. Both RDFs were composed of cellulose and plastics based materials. With increasing temperature from 650℃ to 800℃, concentrations of H2, CO and CH4 increased in both RDFs gasification. Gas yield and gasification efficiency were also improved. The combustible components first increased slightly and then decreased with increasing ER, while gas yield kept constant growth. The optimum ER values for RDF1 and RDF2 were 0.22 and 0.27 respectively for obtaining the highest gasification efficiency. The use of enriched air could improve gasification effectively and lead to higher heating value of the syngas. When oxygen percentage of enriched air was 45%, the maximum low heating values of the syngas for RDF1 and RDF2 were 8.6 MJ·m−3and 9.2 MJ·m−3respectively.%在鼓泡流化床上进行两种垃圾衍生燃料(RDF)的富氧气化试验,考察了RDF的热重特性并分析了气化温度、当量比及富氧浓度对气化特性的影响。结果表明:两种RDF均由纤维素及塑料类组分构成。随着温度由650℃升高至800℃,两种RDF产气的H2、CO及CH4浓度均逐渐增加,产气热值和气化效率同时提高。当量比增大时可燃组分浓度先略有增大后逐渐减小,但气体产率不断增大。RDF1及RDF2分别在当量比为0.22及0.27处达到最佳气化效率。富氧气化可有效改善气化品质,提升合成气热值,富氧浓度为45%时RDF1及RDF2合成气热值均达到最大,分别为8.6 MJ·m−3及9.2 MJ·m−3。

  13. Review of 15 years: high-density low-enriched UMo dispersion fuel development for research reactors in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Berghe, S. van den [SCK.CEN, Nuclear Materials Science Institute (NMS), Boeretan (Belgium); Lemoine, P [Commissariat a l' Eergie Atomique, CEA Saclay, Yvette Cedex (France)

    2014-04-15

    This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R and D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.

  14. Comments on the Joint Proposed Rulemaking to Establish Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards

    Energy Technology Data Exchange (ETDEWEB)

    Wenzel, Tom [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2009-10-27

    Tom Wenzel of Lawrence Berkeley National Laboratory comments on the joint rulemaking to establish greenhouse gas emission and fuel economy standards for light-duty vehicle, specifically on the relationship between vehicle weight and vehicle safety.

  15. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

    Ishibashi, Yoko; Koyama, Jun-ichi; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

  16. Uranium enrichment (a strategy analysis overview)

    International Nuclear Information System (INIS)

    An analysis of available information on enrichment technology, separative work supply and demand, and SWU cost is presented. Estimates of present and future enrichment costs are provided for use in strategy analyses of alternate nuclear fuel cycles and systems. (auth)

  17. Electricity generation and microalgae cultivation in microbial fuel cell using microalgae-enriched anode and bio-cathode

    International Nuclear Information System (INIS)

    Highlights: • Electricity generation and microalgae cultivation was done simultaneously. • Microalgae biomass was used as a substrate at anode. • Freshwater microalgae were grown at cathode. • The maximum power output of 1926 ± 21.4 mW/m2 was achieved. • Microalgae produced biomass up to 1247 ± 52 mg/L. - Abstract: In this study, a microbial fuel cell (MFC) was developed to treat waste, produce electricity and to grow microalgae simultaneously. Dead microalgae biomass (a potential pollution vector in streams) was used as a substrate at anode. CO2 generated at anode was used to grow freshwater microalgae at cathode. The performance of microalgae-fed MFC was compared with acetate-fed MFC. The maximum power density of 1926 ± 21.4 mW/m2 (8.67 ± 0.10 W/m3, at Rext = 100 Ω) and Coulombic efficiency (CE) of 6.3 ± 0.2% were obtained at 2500 mg COD/L of microalgae powder (0.5 g/L). Microalgae captured CO2 (5–14%, v/v) to produce a biomass concentration of 1247 ± 52 mg/L. However, microalgae could not grow in acetate-fed (0.5 g/L) MFC (acetate-control) and without anodic CO2 supplying MFC (CO2-control)

  18. New generation of CASTOR registered casks for high enriched, high burn-up fuel from German NPP

    International Nuclear Information System (INIS)

    Requirements for new cask designs for transport and long-term dry storage of spent fuel assemblies (FA) from LWR-reactors are based on both increased source terms of the LWR FA including MOX FA, as well as the demand for economical optimisation of decommissioning costs by increased cask capacities. For this, cask development is the challenge to create and establish cask designs that can accommodate more FA with higher source terms, each under fixed boundary conditions (i.e. transport requirements and limitations of the power plants as crane loads and/or fixed maximum dimensions). This task has been elaborated by working simultaneously on different development actions each focussed to improve the cask performance. In the following a brief summary will be presented to give an overview which developments and investigations have been and are still will be performed for development and safety analyses of the new CASTOR registered -designs under the main subjects: material investigation and qualification, component tests and verifications, detailed design analysis and not at least design verification

  19. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle for use in establishing ''as low as practicable'' guides: fabrication of light-water reactor fuel from enriched uranium dioxide

    International Nuclear Information System (INIS)

    A cost-benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model enriched-uranium, light-water reactor (LWR) fuel fabrication plant, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as practicable'' in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment equipment is added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Some of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. (U.S.)

  20. Experiments on light water lattices with enriched uranium fuel; Analyse des donnees experimentales sur les reseaux a eau legere et uranium enrichi

    Energy Technology Data Exchange (ETDEWEB)

    Audinet, M. [Societe des Forges et Ateliers du Creusot, 75 - Paris (France); Lamare, J. de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Panossian, J. [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    Experiments a light water lattices with slightly enriched uranium fuel, have been performed at Brookhaven and Bettis Plant Laboratories. The results are studied and compared with simple theories on reactor calculations. By taking into account shadow effects and non Maxwellian neutron spectrum, which are important in this kind of reactors, we have been able to explain the observed results fairly well. We can thus give a constituent set of formulas with which to calculate lattices similar to there we studied. (author) [French] Les resultats d'experiences effectuees aux Laboratoires de Brookbaven et de Bettis Plant, sur des reseaux heterogenes a eau legere et uranium metallique legerement enrichi, sont analyses et confrontes avec les theories simples du calcul de pile. En tenant compte des effets d'interaction et d'echauffement du spectre de neutrons qui sont importants dans ce type de reacteurs, on parvient a rendre compte convenablement des resultats observes. On a ainsi mis au point un formulaire permettant le calcul des reseaux quivpeuvent etre consideres comme assez semblables aux reseaux etudies. (auteur)

  1. Communication dated 30 May 2007 from the Permanent Mission of the United Kingdom of Great Britain and Northern Ireland to the IAEA concerning enrichment bonds - A voluntary scheme for reliable access to nuclear fuel

    International Nuclear Information System (INIS)

    The Secretariat has received a letter dated 30 May 2007 from the Permanent Mission of the United Kingdom of Great Britain and Northern Ireland to the IAEA attaching a UK Non-paper entitled 'Food for Thought: Enrichment Bonds - A Voluntary Scheme for Reliable Access to Nuclear Fuel'. As requested in that letter, the letter and the attachment is now being circulated for the information of all Member States

  2. Is Job Enrichment Really Enriching?

    OpenAIRE

    Robert D. Mohr; Cindy Zoghi

    2006-01-01

    This study uses a survey of Canadian workers with rich, matched data on job characteristics to examine whether “enriched” job design, with features like quality circles, feedback, suggestion programs, and task teams, affects job satisfaction. We identify two competing hypotheses on the relationship between enriched jobs and job satisfaction. The “motivation hypothesis,” implies that enrichment will generally increase satisfaction and the “intensification hypothesis,” implies that enrichment m...

  3. Comments on the Joint Proposed Rulemaking to Establish Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards

    Energy Technology Data Exchange (ETDEWEB)

    Wenzel, Thomas P

    2009-10-27

    I appreciate the opportunity to provide comments on the joint rulemaking to establish greenhouse gas emission and fuel economy standards for light-duty vehicles. My comments are directed at the choice of vehicle footprint as the attribute by which to vary fuel economy and greenhouse gas emission standards, in the interest of protecting vehicle occupants from death or serious injury. I have made several of these points before when commenting on previous NHTSA rulemakings regarding CAFE standards and safety. The comments today are mine alone, and do not necessarily represent the views of the US Department of Energy, Lawrence Berkeley National Laboratory, or the University of California. My comments can be summarized as follows: (1) My updated analysis of casualty risk finds that, after accounting for drivers and crash location, there is a wide range in casualty risk for vehicles with the same weight or footprint. This suggests that reducing vehicle weight or footprint will not necessarily result in increased fatalities or serious injuries. (2) Indeed, the recent safety record of crossover SUVs indicates that weight reduction in this class of vehicles resulted in a reduction in fatality risks. (3) Computer crash simulations can pinpoint the effect of specific design changes on vehicle safety; these analyses are preferable to regression analyses, which rely on historical vehicle designs, and cannot fully isolate the effect of specific design changes, such as weight reduction, on crash outcomes. (4) There is evidence that automakers planned to build more large light trucks in response to the footprint-based light truck CAFE standards. Such an increase in the number of large light trucks on the road may decrease, rather than increase, overall safety.

  4. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

    The paper discusses some aspects of high burnup program implementation at Dukovany NPP and its influence on the fuel cycle costs. Dukovany internal fuel cycle is originally designed as a three years cycle of the Out-In-In fuel reloading patterns. These reloads are not only uneconomical but they additionally increased the radiation load of the reactor pressure vessel due to high neutron leakage typical for Out-In-In loading pattern. To avoid the high neutron leakage from the core a transition to 4-year fuel cycle is started in 1987. The neutron leakage from the core is sequentially decreased by insertion of older fuel assemblies at the core periphery. Other developments in fuel cycle are: 1) increasing of enrichment in control assemblies (3.6% of U-235); 2) improvement in fuel assembly design (reduce the assembly shroud thickness from 2.1 to 1.6 mm); 3) introduction of Zr spacer grid instead of stainless steel; 4) introduction of new type of assembly with profiled enrichment with average value of 3.82%. Due to increased reactivity of the new assemblies the transition to the partial 5-year fuel cycle is required. Typical fuel loading pattern for 3, 3.5, 4 and 5-year cycles are shown in the presented paper. An evaluation of fuel cost is also discussed by using comparative analysis of different fuel cycle options. The analysis shows that introduction of the high burnup program has decrease relative fuel cycle costs

  5. Uranium enrichment

    International Nuclear Information System (INIS)

    GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

  6. Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

    Science.gov (United States)

    Smirnov, A. Yu; Sulaberidze, G. A.; Dudnikov, A. A.; Nevinitsa, V. A.

    2016-09-01

    The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment.

  7. Contribution of CERCA to the US DOE conference on the use of 20% and 45% enriched uranium as fuel for research reactors [contributed by J. Doumerc, CERCA

    International Nuclear Information System (INIS)

    This paper speaks only of prices. Some basic statements can be provided. All the results which have been displayed by Mr. Dewez represent the CERCA work performed within the last twelve months. We have invested in this a little less than million French francs, which is roughly 220 000. As far as prices are concerned, for the time being, we have to compare the prices of a steady state situation, which is represented by a well known process that has been in use for many years, with a transition situation which is the achievement of the same expertise in new, extrapolated fuels. That is why the comparison has to be corrected for the results within the next 2 to 3 years. Obviously, it seems that there are other factors which contribute to the price increases. I think there is a very significant example for this. When you have to introduce a given amount for 235-U in the fuel, either 93 or 20% enrichment, you need in the second case a higher total uranium content, which means that you have to convert two to four times more uranium from UF6 to a uranium compound then from uranium compound to powder. Obviously, you cannot prepare Kg of some product at one price and 250 g of the same product at the same price. Moreover, there is some chance that the new process fabrication will be slightly more difficult to achieve than the previous one. We have observed in the past that progress was continuously improved, until we reached something like a steady state situation. Now alloy yields, for instance, are the same for all manufacturers, except ± 1% depending on the day-to-day events of the manufacturer itself. There is a very good reason to consider that the progress in the yields will be the same for the manufacturer of extrapolated fuels. In that event it is rather easy to foresee what it will be in the near future. In consideration, besides explaining some reasons as to why prices will become higher, there are many reasons of remaining confident with the final result. Because

  8. Parametric study of thermo-mechanical behaviour of 19-element PHWR fuel bundle having AHWR fuel material

    International Nuclear Information System (INIS)

    AHWR Th-LEU of 4.3 weight % 235U enrichment is a fuel design option for its trial irradiation in Indian PHWRs. The important component of this option is the large enhancement in the average discharge burn-up from the core. A parametric study of the 19-element fuel bundle, with natural uranium currently is being used in all operating 220 MWe PHWRs, has been carried out for AHWR Th-LEU fuel material by computer code FUDA MOD2. The important fuel parameters such as fuel temperature, fission gas release, fuel swelling and sheath strain have been analyzed for required fuel performance. With Th-LEU, average discharge burnups of about 25,000 MW-d/TeHE can be achieved. The FUDA code (Fuel Design Analysis code) MOD2 version has been used in the fuel element analysis. The code takes into account the inter-dependence of different parameters like fuel pellet temperatures, pellet expansions, fuel-sheath gap heat transfer, sheath strain and stresses, fission gas release and gas pressures, fuel densification etc. Thermo-mechanical analysis of fuel element having AHWR material is carried out for the bundle power histories reaching up to design burn-up 40000 MWd/TeHE. The resultant parameters such as fuel temperature, sheath plastic strain and fission gas pressure for AHWR fuel element were compared with respective thermo-mechanical parameters for similar fuel bundle element with natural uranium as fuel material. (author)

  9. Comparison of DUPIC fuel composition heterogeneity control methods

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ko, Won Il

    1999-08-01

    A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for the DUPIC fuel which is made of spent pressurized water reactor (PWR) fuels by a dry refabrication process. This study focuses on the reactivity control method which uses either slightly enriched, depleted, or natural uranium to minimize the cost rise effect on the manufacturing of DUPIC fuel, when adjusting the excess reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high reactivity spent PWR fuels, and reactivity control by natural uranium for linear reactivity spent PWR fuels. The results of this study have shown that the reactivity control by slightly enriched and depleted uranium, all the spent PWR fuels can be utilized as the DUPIC fuel and the fraction of fresh uranium feed is 3.4% on an average. For the reactivity control by natural uranium, about 88% of spent PWR fuel can be utilized as the DUPIC fuel when the linear reactivity spent PWR fuels are used, and the amount of natural uranium feed needed to control the DUPIC fuel reactivity is negligible. (author). 13 refs., 16 tabs., 6 figs.

  10. Oxygen enrichment incineration

    International Nuclear Information System (INIS)

    Oxygen enriched combustion technology has recently been used in waste incineration. To apply the oxygen enrichment on alpha-bearing waste incineration, which is being developed, a state-of-an-art review has been performed. The use of oxygen or oxygen-enriched air instead of air in incineration would result in increase of combustion efficiency and capacity, and reduction of off-gas product. Especially, the off-gas could be reduced below a quarter, which might reduce off-gas treatment facilities, and also increase an efficiency of off-gas treatment. However, the use of oxygen might also lead to local overheating and high nitrogen oxides (NOx) formation. To overcome these problems, an application of low NOx oxy-fuel burner and recycling of a part of off-gas to combustion chamber have been suggested

  11. Oxygen enrichment incineration

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Guk; Yang, Hee Chul; Park, Geun Il; Kim, Joon Hyung

    2000-10-01

    Oxygen enriched combustion technology has recently been used in waste incineration. To apply the oxygen enrichment on alpha-bearing waste incineration, which is being developed, a state-of-an-art review has been performed. The use of oxygen or oxygen-enriched air instead of air in incineration would result in increase of combustion efficiency and capacity, and reduction of off-gas product. Especially, the off-gas could be reduced below a quarter, which might reduce off-gas treatment facilities, and also increase an efficiency of off-gas treatment. However, the use of oxygen might also lead to local overheating and high nitrogen oxides (NOx) formation. To overcome these problems, an application of low NOx oxy-fuel burner and recycling of a part of off-gas to combustion chamber have been suggested.

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

    International Nuclear Information System (INIS)

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

  13. The RERTR [Reduced Enrichment Research and Test Reactor] program:

    International Nuclear Information System (INIS)

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) program is described. After a brief summary of the results which the RERTR program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results and new developments which ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U3Si2-Al and U3Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U3Si2-Al fuel at 4.8 g U/cm3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40 % average burnup. Good progress was made in the area of LEU usage for the production of fission 99Mo, and in the coordination of safety evaluations related to LEU conversions of U.S. university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U3Si-Al with 19.75 % enrichment and U3Si2-Al with 45 % enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR program. (Author)

  14. Advances in uranium enrichment processes

    International Nuclear Information System (INIS)

    Advances in gas centrifuges and development of the atomic vapour laser isotope separation process promise substantial reductions in the cost of enriched uranium. The resulting reduction in LWR fuel costs could seriously erode the economic advantage of CANDU, and in combination with LWR design improvements, shortened construction times and increased operational reliability could allow the LWR to overtake CANDU. CANDU's traditional advantages of neutron economy and high reliability may no longer be sufficient - this is the challenge. The responses include: combining neutron economy and dollar economy by optimizing CANDU for slightly enriched uranium fuel; developing cost-reducing improvements in design, manufacture and construction; and reducing the cost of heavy water. Technology is a renewable resource which must be continually applied to a product for it to remain competitive in the decades to come. Such innovation is a prerequisite to Canada increasing her share of the international market for nuclear power stations. The higher burn-up achievable with enriched fuel in CANDU can reduce the fuel cycle costs by 20 to 40 percent for a likely range of costs for yellowcake and separative work. Alternatively, some of the benefits of a higher fissile content can take the form of a cheaper reactor core containing fewer fuel channels and less heavy water, and needing only a single fuelling machine. An opportunity that is linked to this need to introduce an enriched uranium fuel cycle into CANDU is to build an enrichment business in Canada. This could offer greater value added to our uranium exports, security of supply for enriched CANDUs, technological growth in Canada and new employment opportunities. AECL has a study in progress to define this opportunity

  15. Advanced nuclear fuel study for the utilization of carbon-coated

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee Unviersity, Seoul (Korea)

    1998-03-01

    Advanced nuclear fuel design of carbon coated fuel particles(UCO fuel) was suggested to the current PWRs. Nuclear feasibility studying was forformed for the double heterogeneous UCO fuel by CASMO-3. UCO fuel showed nuclear feasibility when they were packed in the Ulchin3/4 fuel assembly. Nuclear safety was evaluated for the UCO fuel by FTC an dMTC, which had enough safety at operating condition. The average fuel temperature compared with conventional oxide fuel at hot full power condition was reduced by 150 deg K, which was caused by high conductivity of carbon matrix. A core design, used UCO fuel, was possible for same forformance with Ulchin3/4. But, UCO fuel enrichment exceed the PWR fuel enrichment limit 5w/o. Cycle length of UCO duel core was shortened by 90 EFPD satisfied with enrichment limit and thermal power. It is not good for using UCO fuel in PWRs in respect of fuel costs. (author). 19 refs., 71 figs., 25 tabs.

  16. Job Enrichment

    Science.gov (United States)

    Sanders, Rick

    1970-01-01

    Job enrichment means giving people more decision-making power, more responsibility, more grasp of the totality of the job, and a sense of their own importance in the company. This article presents evidence of the successful working of this approach (Donnelly Mirrors), and the lack of success with an opposing approach (General Motors). (NL)

  17. The three cylinder Ecotec Compact Engine from Opel with port deactivation - a contribution to reduce the fleet average fuel consumption; Der Dreizylinder Ecotec Compact Motor von Opel mit Kanalabschaltung - ein Beitrag zur Absenkung des Flottenverbrauchs

    Energy Technology Data Exchange (ETDEWEB)

    Grebe, U.D. [Opel (A.) AG, Ruesselsheim (Germany); Kapus, P.E.; Poetscher, P. [AVL List GmbH, Graz (Austria)

    1999-07-01

    The Ecotec Compact Engine, introduced in 1997 by Adam Opel AG, utilizes four valve technology consequently for the reduction of fuel consumption. The introduction of the 1.0 l three cylinder engine in the Opel Corsa resulted in a reduction of fuel consumption of 11% in the European MVEG cycle compared to the 1.2 l two valve engine. This paper describes the application of a port deactivation and a high EGR rate system. Due to the high combustion stability it is possible to apply very high EGR rates of up to 25% in the vehicle. This charge dilution leads to a remarkable dethrottling of the engine at part load. Due to this and additional measures to reduce engine friction fuel consumption in the MVEG cycle could be reduced by additional 10.5% to 5.1 l/100 ml. This engine was also used for the demonstration of a so called '3-liter-car' (90 g CO{sub 2}/km). With the 750 kg concept car 'G90' presented at the 1999 Frankfurt International Motor Show (IAA) it was for the first time possible to approach the 90 g CO{sub 2}/km border with a conventional gasoline engine with port fuel injection. The consequent improvement of the engine while maintaining mixture preparation with port fuel injection leads to a considerable improvement in fuel consumption with acceptable system complexity. In that way an attractive price of the new vehicle for the customer can be realized in combination with very low operating expenses over the lifetime of the vehicle. Only by doing so it is possible to have great influence on the sales-weighted fleet average fuel consumption. (orig.) [German] Der 1997 von der Adam Opel AG vorgestellte Ecotec Compact Motor nutzt die Vierventiltechnik konsequent zur Verbrauchsreduzierung. Der 1,0 l Dreizylindermotor ermoeglichte im Opel Corsa eine Verbrauchsreduzierung im MVEG Testzyklus von annaehernd 11% gegenueber dem Vorgaenger mit 1,2 l Hubraum und Zweiventiltechnik. In diesem Beitrag wird die Anwendung einer Kanalabschaltung und eines Hoch

  18. 40 CFR 80.67 - Compliance on average.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Compliance on average. 80.67 Section...) REGULATION OF FUELS AND FUEL ADDITIVES Reformulated Gasoline § 80.67 Compliance on average. The requirements... with one or more of the requirements of § 80.41 is determined on average (“averaged gasoline”)....

  19. Uranium enrichment

    International Nuclear Information System (INIS)

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector

  20. The Software Design for252Cf Neutron Activation Fuel Rod 235U Enrichment Inspecting Equipment%252Cf中子活化核燃料棒235U富集度检测设备的软件设计

    Institute of Scientific and Technical Information of China (English)

    张雷; 刘明; 马金波

    2013-01-01

    It introduces the software design for 252Cf neutron activation fuel red235U enrichment inspecting equipment.It used multithread technique to control Advantech PCI-1780 counter/timer card,and collect γ-ray signal from the six-path detectors.Process and analyze the collected data can exactly check the actual 235U enrichment and abnormal pellets in the nuclear fuel rods.The software can measure the actual 235U enrichment and judge whether there are abnormal pellets in the nuclear fuel rods accurately,and send customizing messages to PLC which complete automatic sorting,at 6 m/min detection speed.Now the software is used on nondestructive test equipment in Nuclear Fuel Element Factory.%介绍了252Cf中子活化核燃料棒235U富集度检测设备的软件设计,该软件采用多线程技术控制研华PCI-1780采集卡定时采集六路探测器输出的经252Cf中子活化后235U裂变产物的γ射线信号,针对采集数据的特性,进行相应的处理和分析,可以检测出核燃料棒的实际235U富集度以及有无异常芯块.该软件经过实验验证在检测速度为6时,能够准确测量核燃料棒的实际235U富集度值并判断棒中是否混有异常芯块,同时向PLC发送相应信号实现自动分选.目前已应用在核燃料元件厂的核燃料棒235U富集度无损检测设备上.

  1. Uranium enrichment: technology, economics, capacity

    International Nuclear Information System (INIS)

    Large-scale enrichment of uranium has now been carried out for 40 years. While the gaseous diffusion process was the original choice of several countries and continues today to provide the major component of the world production of separative work, the last two decades have witnessed the development of a number of alternative processes for enrichment. These processes, which are being studied and deployed around the world, offer a wide range of technical and economic characteristics which will be useful in assuring adequate capacity to meet projected reactor fuel market needs through the rest of this century at competitive prices. With present uncertainties in future enriched uranium needs, it is apparent that flexibility in the deployment and operation of any enrichment process will be one of the prime considerations for the future. More economical production of separative work not only can have a beneficial impact on reactor fuel costs, but also tends to conserve natural uranium resources. This paper reviews the world scene in the enrichment component of the fuel cycle, including existing or planned commercial-scale facilities and announced R and D efforts on various processes

  2. Fuel cycle data survey

    International Nuclear Information System (INIS)

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

  3. Cycle-by-cycle Variations in a Direct Injection Hydrogen Enriched Compressed Natural Gas Engine Employing EGR at Relative Air-Fuel Ratios.

    Directory of Open Access Journals (Sweden)

    Olalekan Wasiu Saheed

    2014-07-01

    Full Text Available Since the pressure development in a combustion chamber is uniquely related to the combustion process, substantial variations in the combustion process on a cycle-by-cycle basis are occurring. To this end, an experimental study of cycle-by-cycle variation in a direct injection spark ignition engine fueled with natural gas-hydrogen blends combined with exhaust gas recirculation at relative air-fuel ratios was conducted. The impacts of relative air-fuel ratios (i.e. λ = 1.0, 1.2, 1.3 and 1.4 which represent stoichiometric, moderately lean, lean and very lean mixtures respectively, hydrogen fractions and EGR rates were studied. The results showed that increasing the relative air-fuel ratio increases the COVIMEP. The behavior is more pronounced at the larger relative air-fuel ratios. More so, for a specified EGR rate; increasing the hydrogen fractions decreases the maximum COVIMEP value just as increasing in EGR rates increases the maximum COVIMEP value. (i.e. When percentage EGR rates is increased from 0% to 17% and 20% respectively. The maximum COVIMEP value increases from 6.25% to 6.56% and 8.30% respectively. Since the introduction of hydrogen gas reduces the cycle-by-cycle combustion variation in engine cylinder; thus it can be concluded that addition of hydrogen into direct injection compressed natural gas engine employing EGR at various relative air-fuel ratios is a viable approach to obtain an improved combustion quality which correspond to lower coefficient of variation in imep, (COVIMEP in a direct injection compressed natural gas engine employing EGR at relative air-fuel ratios.

  4. Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

    International Nuclear Information System (INIS)

    Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U3O8 (Megawatt Years Thermal per Short Ton of U3O8). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U3O8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

  5. Final report on the irradiation testing and post-irradiation examination of low enriched U3O8-Al and UAlx-Al fuel elements by the Netherlands Energy Research Foundation (ECN)

    International Nuclear Information System (INIS)

    Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and subjected to post-irradiation examination. Two of the elements contain UAlx-Al and two contain U3O8-Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the PIE results is given in this report. (author). 5 refs, 31 figs, 3 tabs

  6. Contribution to the study of the evolution of nuclear fuel composition in PWR type reactors. Reactor cores in three and four regions of enrichment

    International Nuclear Information System (INIS)

    The main methods for calculations of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Plant are presented and discussed. (author)

  7. Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

    OpenAIRE

    Graven, H. D.; Gruber, N.

    2011-01-01

    The 14C-free fossil carbon added to atmospheric CO2 by combustion dilutes the atmospheric 14C/C ratio (Δ14C), potentially providing a means to verify fossil CO2 emissions calculated using economic inventories. However, sources of 14C from nuclear power generation and spent fuel reprocessing can counteract this dilution and may bias 14C/C-based estimates of fossil fuel-derived CO2 if these nuclear influences are not correctly accounted for. Previous studies have examined nuclear influenc...

  8. Nuclear fuel

    International Nuclear Information System (INIS)

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts. (Kako, I.)

  9. Neutron resonance averaging

    International Nuclear Information System (INIS)

    The principles of resonance averaging as applied to neutron capture reactions are described. Several illustrations of resonance averaging to problems of nuclear structure and the distribution of radiative strength in nuclei are provided. 30 refs., 12 figs

  10. Field Measurement of Am241 and Total Uranium at a Mixed Oxide Fuel Facility with Variable Uranium Enrichments Ranging from 0.3% to 97% U235

    Energy Technology Data Exchange (ETDEWEB)

    Conway, K. C.

    2002-02-28

    The uranium and transuranic content of site soils and building rubble can be accurately measured using a NaI(Tl) well counter, without significant soil preparation. Accurate measurements of total uranium in uranium-transuranic mixtures can be made, despite a wide range (0.3% to 97%) of uranium enrichment, sample mass, and activity concentrations. The appropriate uranium scaling factors needed to include the undetected uranium isotopes, particularly U 234 can be readily determined on a sample by sample basis as a part of the field analysis, by comparing the relative response of the U 235 186 keV peak versus the K shell X rays of U 238 , U 235, and their immediate ingrowth daughters. The ratio of the two results is a sensitive and accurate predictor of the uranium enrichment and scaling factors. The case study will illustrate how NaI(Tl) gamma spectrometry was used to provide rapid turnaround uranium and transuranic activity levels for soil and building rubble with sample by sample determination of the appropriate scaling factor to include the U234 and Uranium238 content.

  11. Averaging anisotropic cosmologies

    International Nuclear Information System (INIS)

    We examine the effects of spatial inhomogeneities on irrotational anisotropic cosmologies by looking at the average properties of anisotropic pressure-free models. Adopting the Buchert scheme, we recast the averaged scalar equations in Bianchi-type form and close the standard system by introducing a propagation formula for the average shear magnitude. We then investigate the evolution of anisotropic average vacuum models and those filled with pressureless matter. In the latter case we show that the backreaction effects can modify the familiar Kasner-like singularity and potentially remove Mixmaster-type oscillations. The presence of nonzero average shear in our equations also allows us to examine the constraints that a phase of backreaction-driven accelerated expansion might put on the anisotropy of the averaged domain. We close by assessing the status of these and other attempts to define and calculate 'average' spacetime behaviour in general relativity

  12. Radiological and nuclear safety aspects in the fabrication of 1.8% enriched U O{sub 2} fuel rods for the RA-8 critical facility; Aspectos de seguridad radiologica y nuclear en la fabricacion de barras combustibles, con U O{sub 2} enriquecido al 1.8%, para la facilidad critica RA-8

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Hugo; Becarra, Fabian; Herrero, Jorge; Luna, Manuel; Perez, Aldo [Comision Nacional de Energia Atomica, (Argentina). Centro Atomico Constituyentes

    1997-10-01

    The neutronic behavioral study of the fuel for the future nuclear reactor CAREM required to mount critical facility with 1.8% enriched U O{sub 2} fuel rods. The present work describes the various operation and production processes, the safety and radioprotection systems, the administrative procedures and the associated radiological controls. Also, the results obtained in the area and personal monitoring and waste generation are detailed. (author). 10 refs., 4 figs., 1 tab.

  13. The ORR Whole-Core LEU Fuel Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L.

    1990-01-01

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs.

  14. Proliferation resistance and energy security advantages of a thorium-uranium dioxide once-through fuel cycle for light water reactors

    International Nuclear Information System (INIS)

    This study analyzes whether spent light reactor (LWR) thorium-uranium dioxide fuel poses a significantly lower risk for nuclear weapon proliferation than spent uranium-dioxide fuel, based on the isotopic composition of the contained uranium and plutonium. Mixed Th/U fuel with an initial enrichment of 19.5% U235 can achieve an average burnup of 70,000 MWd/tHM in a PWR using 30% UO2 and 70% ThO2. To get the equivalent burnup, LEU fuel requires an initial enrichment of 8.0% U235. Two computer codes, MCNP and ORIGEN2, are used to perform the depletion calculation. The spent mixed thorium-uranium dioxide fuel discharged from a pressurized-water reactor has a plutonium isotopic composition and higher decay heat production per kilogram of plutonium more proliferation resistant than spent low enriched uranium dioxide fuel, while significantly reducing the quantity of plutonium produced. The U233 + U235 mixture in spent thorium-uranium fuel is low enriched and contaminated with gamma-emitting U232. With respect to energy security, the introduction of a thorium-uranium fuel cycle could reduce concern over uranium fuel supply of a resource-poor nation since thorium reserve is much larger, compared to fuel cycles using 4.5% LEU, while its uranium saving is almost equivalent to plutonium recycling. Overall, spent thorium-uranium fuel appears significantly more proliferation resistant in terms of the weapons-usability of the contained fissile material than spent low enriched uranium fuel, although use of 19.5% enriched uranium in fresh fuel would facilitate production of weapons-grade uranium at a higher rate in countries with clandestine enrichment facilities. (S.Y.)

  15. The possible use of cermet fuel in the DIDO and PLUTO heavy-water research reactors

    International Nuclear Information System (INIS)

    International restrictions on the supply of highly enriched uranium have resulted in the requirement to fuel research reactors with a lower-enrichment uranium fuel. A study has been made of the feasibility of using low-enrichment fuels of a new type in the DIDO and PLUTO reactors. This work has been done as a contribution to the studies currently being carried out internationally on the implications of using lower-enrichment fuels in heavy-water-moderated research reactors. The uranium content of the U/Al alloy at present used cannot be increased sufficiently to maintain the requisite U235 content without undesirable effects on the physical properties of the alloy. A different type of fuel will therefore be required to maintain the desired nuclear characteristics. A possible solution to the problem is the use of a cermet (U3O8/Al) fuel material. Cermet fuel has poorer thermal conductivity than metallic fuel, and may also contain particles of the ceramic of a size that approaches the total thickness of the cermet core. We therefore have to consider both the average temperature of the centre of the fuel and whether large particles of the ceramic may be significantly hotter than the average. This paper describes a preliminary study of the feasibility of this concept from the heat-transfer and safety viewpoints. Calculations have been made for a cermet of 20%-enrichment 2.3g U/cm3, used in a high-power element in a DIDO-type reactor. To accommodate the cermet, the cladding has been reduced in thickness to 0.318mm (0.0125 in) the core increasing to 1.044mm, but the fuel geometry is otherwise unchanged. It is concluded that from the heat-transfer viewpoint there is no problem during normal operation or the maximum credible power transient in these reactors. (author). 10 refs, 6 figs, 2 tabs

  16. Investigation of deuterium cross section data by integral testing: ZED-2 measurements of high-enriched uranium fuel substituted into a natural uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Atfield, J.E.; Kozier, K.S.; Roubtsov, D.; Zeller, M.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    Historical ZED-2 measurements of an HEU fuel rod substituted into a lattice of NU rods were analysed to determine their reactivity sensitivity to differences between the neutron elastic scattering cross-sections of deuterium from different evaluated nuclear data libraries. The differences in the deuterium nuclear data concern the angular probability distribution at neutron energies below 3.2 MeV. These ZED-2 experiments were selected due to the presence of HEU fuel in D{sub 2}O, since analyses of other critical experiments involving solutions of HEU fluoride in D{sub 2}O show substantial sensitivity (~10 mk) to these differences in the deuterium nuclear data. This analysis shows that the existing ZED-2 HEU experiments are insufficiently sensitive to resolve the discrepancy between the different deuterium data libraries. Further analysis of hypothetical configurations with high sensitivity shows that the sensitivity to the angular probability distribution of deuterium is strongly correlated with the leakage of fast neutrons, and it is recommended that further experiments to address this deuterium nuclear data issue be designed/evaluated to maximize this quantity. (author)

  17. Design of Hydrogen Enriched Compressed Natural Gas Engine Fuel System and Test Research%天然气掺氢发动机燃气供给系统设计与试验研究

    Institute of Scientific and Technical Information of China (English)

    张朝山; 熊树生; 任晓帅; 姚红; 徐进; 谢莲; 刘震涛

    2012-01-01

    提出了将滑动弧电解制氢装置应用到天然气发动机中,通过电解天然气制氢,轻松实现天然气(CNG)发动机到天然气掺氢(HCNG)发动机的改装.通过自制装置,进行了过量空气系数和点火提前角与燃用不同掺氢比例的HCNG对发动机排放特性影响的试验研究.结果表明,发动机燃用HCNG,其HC和CO的排放都减少,NOx排放量增加,但随着过量空气系数的增加或点火提前角的减少,NOx排放会大大减少,排放性能得到优化.同时进行了体积掺氢比20%的HCNG和纯CNG外特性对比试验研究,结果表明,相比纯CNG,燃用掺氢20% HCNG后,其动力性变化不大,燃料消耗率却相应的减少,经济性得到改善.%This paper put forward a device installed into the fuel system,which used the sliding electric arc to electrolyze the natural gas to make hydrogen and then blend them into the fuel pipe for final combustion. We could re-equip the compressed natural gas (CNG) engine to hydrogen enriched compressed natural gas (HCNG) engine easily. The test research of the engine emissions characteristics using different hydrogen-CNG ratios was conducted when the excess air ratios and the spark advance angles were different. The results show that HC and CO emissions of engine fueled with HCNG reduce when NOx emissions increase. The NOx emissions are reduced greatly with the increase of the excess air ratio or the decrease of the spark advance angle. Comparative experiments of the performance characteristics of engine burned with HCNG whose volume hydrogen-CNG ratio was 20% and CNG were conducted under wide open throttle operating conditions. The results show that the torque output is unchanged when 20% HCNG is burned compared with CNG engine,but the fuel consumption is reduced and the fuel economy is improved.

  18. A Study on Securing Uranium Enrichment Supply Assurance in Korea

    International Nuclear Information System (INIS)

    As a non-nuclear-weapon state without sensitive fuel cycle facilities, Korea is keen about how to secure nuclear fuel supply assurance, especially uranium enrichment for their sustainable nuclear power. However when sensitive nuclear fuel cycle activity such as uranium enrichment is pursued by a national approach, neighboring countries and the world would show concerns about possibility of its proliferation. Therefore, it is critical to allay proliferation concerns by the international community if a country wants to have uranium enrichment capability for its fuel supply assurance. This study describes how to secure uranium enrichment supply assurance in Korea. From the aspect of securing uranium enrichment supply assurance, there are a few conceivable options in Korea from relying on the existing market with buying ownership to establishing a domestic enrichment capability

  19. Conceptual design of KALIMER uranium metallic fueled core

    International Nuclear Information System (INIS)

    As a part of the core design development of KALIMER(150 MWe), the KALIMER core design which uses U-Zr binary fuel not in excess of 20% enrichment was performed. Starting from the former uranium metallic fueled core design, a more economic and safer equilibrium core design was first established based on extensive researches for the possible enrichment gains over various design options and in-core fuel management strategies. Further optimization to extend fuel discharge burnup has been achieved by employing strategic loading schemes for initial and transition cycles to reach the equilibrium cycle early. The core performance analysis based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and core average discharge burnup of 61.6 MWD/kg. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. When comparing with conventional plutonium metallic fueled cores of the same power level, the present KALIMER uranium metallic fueled core has an increased physical core size to meet the enrichment restriction, and, as a result, a lower power density to realize the minimum one-year cycle operation. The KALIMER uranium metallic fueled core characterized by its negative sodium void reactivity and low power density can be operated with maximizing its core safety characteristics as a first generation LMR. The present uranium metallic fueled core allows an easy replacement with different fuel compositions by its demands, with the accumulation of operation experience and design data verification. (author). 34 refs., 34 tabs., 12 figs.

  20. Reactivity measurements on an experimental assembly of 4.31 wt % 235U enriched UO2 fuel rods arranged in a shipping cask geometry

    International Nuclear Information System (INIS)

    A research program was initiated for the US Department of Energy (DOE) Sandia National Laboratory Transportation Systems Development Department in 1982 to provide benchmark type experimental criticality data in support of the design and safe operations of nuclear fuel transportation systems. The overall objective of the program is to identify and provide the experimental data needed to form a consistent, firm, and complete data base for verifying calculational models used in the criticality analyses of nuclear transport and related systems. A report, PNL-6205, issued in June 1988 (Bierman 1988) covered measurement results obtained from a series of experimental assemblies (TIC-1, 2, 3 and 4) involving neutron flux traps. The results obtained on a fifth experimental assembly (TIC-5), modeled after a calculational problem of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) Working Group, are covered in this report. 10 refs., 10 figs., 7 tabs

  1. Average-energy games

    OpenAIRE

    Bouyer, Patricia; Markey, Nicolas; Randour, Mickael; Larsen, Kim G.; Laursen, Simon

    2015-01-01

    Two-player quantitative zero-sum games provide a natural framework to synthesize controllers with performance guarantees for reactive systems within an uncontrollable environment. Classical settings include mean-payoff games, where the objective is to optimize the long-run average gain per action, and energy games, where the system has to avoid running out of energy. We study average-energy games, where the goal is to optimize the long-run average of the accumulated energy. We show that this ...

  2. Averaged extreme regression quantile

    OpenAIRE

    Jureckova, Jana

    2015-01-01

    Various events in the nature, economics and in other areas force us to combine the study of extremes with regression and other methods. A useful tool for reducing the role of nuisance regression, while we are interested in the shape or tails of the basic distribution, is provided by the averaged regression quantile and namely by the average extreme regression quantile. Both are weighted means of regression quantile components, with weights depending on the regressors. Our primary interest is ...

  3. On the Averaging Principle

    OpenAIRE

    Fibich, Gadi; Gavious, Arieh; Solan, Eilon

    2012-01-01

    Typically, models with a heterogeneous property are considerably harder to analyze than the corresponding homogeneous models, in which the heterogeneous property is replaced with its average value. In this study we show that any outcome of a heterogeneous model that satisfies the two properties of differentiability and interchangibility is O(\\epsilon^2) equivalent to the outcome of the corresponding homogeneous model, where \\epsilon is the level of heterogeneity. We then use this averaging pr...

  4. Average Angular Velocity

    OpenAIRE

    Van Essen, H.

    2004-01-01

    This paper addresses the problem of the separation of rotational and internal motion. It introduces the concept of average angular velocity as the moment of inertia weighted average of particle angular velocities. It extends and elucidates the concept of Jellinek and Li (1989) of separation of the energy of overall rotation in an arbitrary (non-linear) $N$-particle system. It generalizes the so called Koenig's theorem on the two parts of the kinetic energy (center of mass plus internal) to th...

  5. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

    1999-01-01

    In this article two common approaches to averaging rotations are compared to a more advanced approach based on a Riemannian metric. Very offten the barycenter of the quaternions or matrices that represent the rotations are used as an estimate of the mean. These methods neglect that rotations belong...... approximations to the Riemannian metric, and that the subsequent corrections are inherient in the least squares estimation. Keywords: averaging rotations, Riemannian metric, matrix, quaternion...

  6. Averaging anisotropic cosmologies

    CERN Document Server

    Barrow, J D; Barrow, John D.; Tsagas, Christos G.

    2006-01-01

    We examine the effects of spatial inhomogeneities on irrotational anisotropic cosmologies by looking at the average properties of pressure-free Bianchi-type models. Adopting the Buchert averaging scheme, we identify the kinematic backreaction effects by focussing on spacetimes with zero or isotropic spatial curvature. This allows us to close the system of the standard scalar formulae with a propagation equation for the shear magnitude. We find no change in the already known conditions for accelerated expansion. The backreaction terms are expressed as algebraic relations between the mean-square fluctuations of the models' irreducible kinematical variables. Based on these we investigate the early evolution of averaged vacuum Bianchi type $I$ universes and those filled with pressureless matter. In the latter case we show that the backreaction effects can modify the familiar Kasner-like singularity and potentially remove Mixmaster-type oscillations. We also discuss the possibility of accelerated expansion due to ...

  7. Composition heterogeneity analysis for DUPIC fuel (I) - statistical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    1999-08-01

    The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: the fissile content adjustment, the reactivity control by slightly enriched and deplete uranium, and the reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF) are close to those of the core that has single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are less than 0.6, 1.5, and 0.8 %, respectively, including the uncertainty of the group-average fuel property. This study has shown that the three DUPIC fuel options reduces the composition heterogeneity effectively and the zone power control system has a sufficient margin to adjust the perturbations caused by the fuel composition heterogeneity. (author). 15 refs., 10 tabs., 28 figs.

  8. Possibility of Different Fuel Cycles Usage in GT-MHR

    International Nuclear Information System (INIS)

    The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis re-indicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another. (authors)

  9. Composition heterogeneity analysis for DUPIC fuel(I) - Statistical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: the fissile content adjustment, the reactivity control by slightly enriched and depleted uranium, and the reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF) are close to those of the core that has single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are less than 0.6, 1.5 and 0.8%, respectively, including the uncertainty of the group-average fuel property. This study has shown that the three DUPIC fuel options reduces the composition heterogeneity effectively and the zone power control system has a sufficient margin to adjust the perturbations cased by the fuel composition heterogeneity. 15 refs., 28 figs.,10 tabs. (Author)

  10. Average Angular Velocity

    CERN Document Server

    Essén, H

    2003-01-01

    This paper addresses the problem of the separation of rotational and internal motion. It introduces the concept of average angular velocity as the moment of inertia weighted average of particle angular velocities. It extends and elucidates the concept of Jellinek and Li (1989) of separation of the energy of overall rotation in an arbitrary (non-linear) $N$-particle system. It generalizes the so called Koenig's theorem on the two parts of the kinetic energy (center of mass plus internal) to three parts: center of mass, rotational, plus the remaining internal energy relative to an optimally translating and rotating frame.

  11. On sparsity averaging

    CERN Document Server

    Carrillo, Rafael E; Wiaux, Yves

    2013-01-01

    Recent developments in Carrillo et al. (2012) and Carrillo et al. (2013) introduced a novel regularization method for compressive imaging in the context of compressed sensing with coherent redundant dictionaries. The approach relies on the observation that natural images exhibit strong average sparsity over multiple coherent frames. The associated reconstruction algorithm, based on an analysis prior and a reweighted $\\ell_1$ scheme, is dubbed Sparsity Averaging Reweighted Analysis (SARA). We review these advances and extend associated simulations establishing the superiority of SARA to regularization methods based on sparsity in a single frame, for a generic spread spectrum acquisition and for a Fourier acquisition of particular interest in radio astronomy.

  12. The RERTR (Reduced Enrichment Research and Test Reactor) Program: Progress and plans

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission /sup 99/Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U/sub 3/Si-Al with 19.75% enrichment and U/sub 3/Si/sub 2/-Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program.

  13. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

    2001-01-01

    In this paper two common approaches to averaging rotations are compared to a more advanced approach based on a Riemannian metric. Very often the barycenter of the quaternions or matrices that represent the rotations are used as an estimate of the mean. These methods neglect that rotations belong...

  14. Analysis of fuel options in TRIGA reactor

    International Nuclear Information System (INIS)

    In this paper, nuclear characteristics of TRIGA Mark-III has been analyzed in detail for six different fuel options. Presently, 70w/o enriched FLIP fuels are adopted for TRIGA core to improve fuel lifetime. However, such highly enriched fuels are not easily obtained due to nonproliferation treaty. This research examines the possible substitution for FLIP fuels with high density fuels without reducing the nuclear performance. This work will provide long-time plan for TRIGA operation (author)

  15. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  16. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

  17. The averaging principle

    OpenAIRE

    Fibich, Gadi; Gavious, Arieh; Solan, Eilon

    2012-01-01

    Typically, models with a heterogeneous property are considerably harder to analyze than the corresponding homogeneous models, in which the heterogeneous property is replaced with its average value. In this study we show that any outcome of a heterogeneous model that satisfies the two properties of \\emph{differentiability} and \\emph{interchangibility}, is $O(\\epsilon^2)$ equivalent to the outcome of the corresponding homogeneous model, where $\\epsilon$ is the level of heterogeneity. We then us...

  18. Reassembling Procedure of the Fuel Assemblies for the Nuclear Power Ship ''Mutsu''

    International Nuclear Information System (INIS)

    Japan's first voyage utilized by nuclear power was made by the nuclear powered ship ''Mutsu'' in 1990. After a research voyage in 1992, decommissioning work of the nuclear reactor for ''Mutsu'' was started to change it from the nuclear power ship to an ordinary power ship. Thirty-four irradiated fuel assemblies of ''Mutsu'' were removed from the reactor and transported to the Reactor Fuel Examination Facility (RFEF) in Nuclear Science Research Institute (NSRI) of Japan Atomic Energy Agency (JAEA). ''Mutsu'' fuel assemblies were loaded into a hot cell of RFEF using the roof gate as the top loading procedure. After the reliability confirmation tests, fuel assemblies were reassembled for reprocessing. To perform the reliability confirmation tests and reassembling, new devices were developed and installed in the hot cells, ''Fuel assembly transportation device'' for transporting the fuel assemblies between the hot cells, ''Upper nozzle cutting device'' for removing the upper nozzle from the fuel assembly, ''Fuel rod drawing device'' for drawing a fuel rod from the fuel assembly and so on. Thirty-four fuel assemblies were reassembled as six PWR type fuel assemblies in order to adjust the acceptable specifications of the reprocessing plant in JAEA: the shape of fuel assembly is the same as the PWR type commercial reactor fuel and the average enrichment of uranium in the assembly is under 4.0%. This paper reports the reassembling techniques of the ''Mutsu'' irradiated fuel assemblies for reprocessing. (author)

  19. Conceptual design report of the SMART fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The SMART fuel rod is based on 17 x 17 KOFA(Korea Fuel Assembly) fuel rod of the 950MWe pressurize water reactor. The fuel stack length of the KOFA is 3658mm, otherwise SMART fuel rod stack length is 2000mm. The fuel rod contains UO{sub 2} pellets with the enrichment of 4.95%. All the fuel in core will be replaced every 35 months. The average LHGR of the fuel rod is 120 W/cm, commercial PWR is 178 W/cm, SMART LHGR is lower about 31% than commercial PWR. The core inlet and outlet temperature of coolant are respectively 270 deg C and 310 deg C, commercial PWR are respectively 291.6 deg C and 326.8 deg C, SMART inlet and outlet temperature is lower averaged 19.2 deg C than commercial PWR. The coolant use mixed soluble ammonia in high purity water and boron is not in. The general performance of the fuel rod UO{sub 2} pellet has been already verified through the sufficient burnup (60,000 MWd/MTU-rod avg.) experience as the rods of same design in commercial PWR's. But cladding corrosion is required the further verification. (author). 13 refs., 3 figs., 8 tabs.

  20. Beyond Job Enrichment to Employment Enrichment

    Science.gov (United States)

    Werther, William B., Jr.

    1975-01-01

    Employment enrichment views the total work environment confronting employees as a system consisting of two overlapping areas: worker-job and worker-organization subsystems. Job enrichment has improved the worker-job subsystem. The focus of this article is on methods of improving the worker-organization relationship. (Author/JB)

  1. 垃圾衍生燃料富氧燃烧污染物排放特性%Pollutants emission characteristics of refuse derived fuel in oxygen-enriched combustion

    Institute of Scientific and Technical Information of China (English)

    李延吉; 姜璐; 赵宁; 李玉龙; 李润东; 池涌

    2013-01-01

    Combustion and emission characteristics of refuse derived fuel ( RDF ) were experimental studied in a tubular high temperature furnace. The results showed that; 1 The increase of plastic proportion in RDF and oxygen concentration in combustion air lead the increase of NOX emission; but under the pure oxygen environment of RDF combustion, NOx emission is greatly reduced; 2 SO2 concentration increases with increasing plastic proportion in RDF and oxygen concentration in combustion air,; 3 Oxygen concentration in combustion air does not have significance effect on CO emission when it is less than 80% in oxygen-enriched combustion,. In pure oxygen combustion, increasing the plastic ratio can reduce the CO emissions; CO emissions tended to decrease with increasing oxygen concentration, CO emissions is minimum in pure oxygen condition; 4 The emissions of NOx, SO2 and CO are lower than national standard, which indicated that the RDF combustion with oxygen enrichment is beneficial for the c pollutant emissions reduction.%为了解决城市生活垃圾直接焚烧产生的二次污染问题,将城市生活垃圾制成垃圾衍生燃料(RDF),在高温管式炉内进行富氧燃烧污染物排放特性研究.结果表明:塑料比例增加,燃烧过程中NOx浓度增大;氧浓度增加,NOx浓度增大;但纯氧条件下RDF燃烧,NOx浓度大大降低.塑料比例增加,燃烧过程中SO2浓度增大;氧浓度增加,SO2浓度降低.当氧浓度为80%时,CO浓度相差不大,纯氧时,塑料比例增大,CO浓度减小;氧浓度增大,CO浓度呈减小趋势,纯氧时CO浓度最小.NOx、SO2、CO的浓度均低于国家标准,说明RDF富氧燃烧有利于降低污染物排放浓度.

  2. Negative Average Preference Utilitarianism

    Directory of Open Access Journals (Sweden)

    Roger Chao

    2012-03-01

    Full Text Available For many philosophers working in the area of Population Ethics, it seems that either they have to confront the Repugnant Conclusion (where they are forced to the conclusion of creating massive amounts of lives barely worth living, or they have to confront the Non-Identity Problem (where no one is seemingly harmed as their existence is dependent on the “harmful” event that took place. To them it seems there is no escape, they either have to face one problem or the other. However, there is a way around this, allowing us to escape the Repugnant Conclusion, by using what I will call Negative Average Preference Utilitarianism (NAPU – which though similar to anti-frustrationism, has some important differences in practice. Current “positive” forms of utilitarianism have struggled to deal with the Repugnant Conclusion, as their theory actually entails this conclusion; however, it seems that a form of Negative Average Preference Utilitarianism (NAPU easily escapes this dilemma (it never even arises within it.

  3. Fuel behavior comparison for a research reactor

    Science.gov (United States)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  4. Project and supply agreement. The text of the agreement of 15 January 1993 between the International Atomic Energy Agency and the Government of the Republic of Indonesia and the Government of the United States of America concerning the transfer of enriched uranium for materials test reactor fuel development

    International Nuclear Information System (INIS)

    The text of the Project and Supply Agreement, which was approved by the Agency's Board of Governors on 4 December 1992 and concluded on 15 January 1993 between the Agency and the Governments of the Republic of Indonesia and the United States of America for the transfer of enriched uranium for materials test reactor fuel development is reproduced herein for the information of all Members. The agreement entered into force on 15 January 1993, pursuant to Article XII.1

  5. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  6. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

  7. DOE enrichment plant hums ahead

    International Nuclear Information System (INIS)

    The Department of Energy's $10-billion gas centrifuge uranium enrichment plant, after three years of construction, is rising on schedule near Piketon, Ohio. A detailed conceptual design, smart management, liberal design fees, hungry contractors and cooperative unions are combining to get the job done. One reason for completing the task is that this will be a far more efficient process - 135 MW will be required to operate the centrifuge plant vs more than 2100 MW to produce the same amount of fuel at the mile-square diffusion plant near Portsmouth, Ohio

  8. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    International Nuclear Information System (INIS)

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  9. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  10. RERTR-2004: International meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Book of abstracts

    International Nuclear Information System (INIS)

    Oral and poster presentations of the Meeting covered the following topics: National and international programs related to Reduced Enrichment for Research and Test Reactors (RERTR); development of new fuel types, testing, fabrication, modelling; studies of reactor cores conversion from highly enriched to low enriched fuel, including licensing; new and converted reactors; spent fuel management including storage and transportation; production of Molybdenum 99 under converted core conditions

  11. Feasibility Study on AFR-100 Fuel Conversion from Uranium-based Fuel to Thorium-based Fuel

    International Nuclear Information System (INIS)

    The feasibility study of converting a fast reactor from uranium-based fuel to thorium-based fuel was studied using the 100 MWe Advanced Fast Reactor (AFR-100). Several fuel conversion scenarios were envisioned in this study. The first scenario is a progressive fuel conversion without fissile support. It consists of progressively replacing the burnt uranium-based fuel with pure thorium-based fuel without fissile material addition. This was found to be impractical because the low excess reactivity of the uranium-fuelled AFR-100 core, resulting in an extremely short cycle length even when only a few assemblies are replaced. A second scenario consists in operating the reference LEU fuelled AFR-100 core for 24 years and then replacing one fuel batch out of four every 7.04 years with thorium-based fuel mixed with transuranics. The transuranics weight fraction required during the transition period is identical to that required at equilibrium and is equal to 18.6%. The original uranium-based fuel is discharged with an average burnup of 120 GWd/t and the Th-TRU fuel with an average burnup of 101 GWd/t. The thermal-hydraulic and passive safety performances of this core are similar to those of the reference AFR-100 design. However, Th-TRU fuel fabrication and performance needs to be demonstrated and TRU separation from the LWR used nuclear fuel is necessary. The third scenario proposed consists of replacing the whole AFR-100 core with fuel assemblies made of several thorium and 20% enriched LEU layers. The mode of operation is similar to that of the reference AFR-100 core with the exception of the cycle length which needs to be reduced from 30 to 18 years. The average LEU and thorium discharge burnups are 79 GWd/t and 23 GWd/t, respectively. The major benefit of this approach is the improved inherent safety of the reactor due to the reduced coolant void worth. (author)

  12. TRIGA-LEU fuel immediately available for substitution in plate-type research reactors up to 15MW

    International Nuclear Information System (INIS)

    New 20% enriched TRIGA type fuel has been developed to replace the TRIGA 70% and 90% enriched fuel and to replace highly enriched uranium plate-type fuel. At present, production elements and undergoing in-pile demonstration testing. The new fuel contains the characteristics of the U-ZrH fuel and can be normally used with the existing core structure. (author)

  13. Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2011-12-01

    The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

  14. Enrichment: Present and projected future supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Lenders, Maurice [URENCO, Buckinghamshire (United Kingdom)

    2009-04-15

    Long term fuel cycle contracts provide reliable supply at predictable cost. By 2015 all operating enrichment capacity may be based on centrifuge. Enrichment capacity expansion will be modular and adjusted to meet demand in a competitive market. Two primary sources of technology (ETC or Russia) can provide all required capacity worldwide. Sufficient enrichment capacity can be installed on time to meet forecast SWU demand for existing and new NPP worldwide.

  15. U.S. forms uranium enrichment corporation

    International Nuclear Information System (INIS)

    After almost 40 years of operation, the federal government is withdrawing from the uranium enrichment business. On July 1, the Department of Energy turned over to a new government-owned entity--the US Enrichment Corp. (USEC)--both the DOE enrichment plants at Paducah, Ky., and Portsmouth, Ohio, and domestic and international marketing of enriched uranium from them. Pushed by the inability of DOE's enrichment operations to meet foreign competition, Congress established USEC under the National Energy Policy Act of 1992, envisioning the new corporation as the first step to full privatization. With gross revenues of $1.5 billion in fiscal 1992, USEC would rank 275th on the Fortune 500 list of top US companies. USEC will lease from DOE the Paducah and Portsmouth facilities, built in the early 1950s, which use the gaseous diffusion process for uranium enrichment. USEC's stock is held by the US Treasury, to which it will pay annual dividends. Martin Marietta Energy Systems, which has operated Paducah since 1984 and Portsmouth since 1986 for DOE, will continue to operate both plants for USEC. Closing one of the two facilities will be studied, especially in light of a 40% world surplus of capacity over demand. USEC also will consider other nuclear-fuel-related ventures. USEC will produce only low-enriched uranium, not weapons-grade material. Indeed, USEC will implement a contract now being completed under which the US will purchase weapons-grade uranium from dismantled Russian nuclear weapons and convert it into low-enriched uranium for power reactor fuel

  16. Development of neural network for analysis of local power distributions in BWR fuel bundles

    International Nuclear Information System (INIS)

    A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

  17. Dissociating Averageness and Attractiveness: Attractive Faces Are Not Always Average

    Science.gov (United States)

    DeBruine, Lisa M.; Jones, Benedict C.; Unger, Layla; Little, Anthony C.; Feinberg, David R.

    2007-01-01

    Although the averageness hypothesis of facial attractiveness proposes that the attractiveness of faces is mostly a consequence of their averageness, 1 study has shown that caricaturing highly attractive faces makes them mathematically less average but more attractive. Here the authors systematically test the averageness hypothesis in 5 experiments…

  18. Profile of World Uranium Enrichment Programs - 2007

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    2007-11-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring weapons grade fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, while HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use in fuel for nuclear reactors. However, the same equipment used to produce LEU for nuclear fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is only enriched to LEU, no undeclared LEU is produced, and no uranium is enriched to HEU or secretly diverted. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity, but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 53 million kg-separative work units (SWU) per year, with 22 million in gaseous diffusion and 31 million in gas centrifuge plants. Another 23 million SWU/year of capacity are under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique

  19. An analysis of LEU fuel behavior as compared to HEU fuel in the 14 MW TRIGA SSR reactor

    International Nuclear Information System (INIS)

    The paper presents an analysis of the behavior and properties of the fuel loading the 14 MW TRIGA research reactor. Comparison are made between the original highly enriched fuel and the slightly enriched fuel which is loading the reactor at present. Both the highly enriched and the slightly enriched fuels have the same physical and thermal properties but different nuclear properties. Thermal hydraulic analysis of the transient regime behavior of reactivity insertion type was effected, evidencing the different behavior of the fuels with different enrichments. (authors)

  20. Reduced Enrichment for Research and Test Reactors. Proceedings of the XVIII international meeting

    International Nuclear Information System (INIS)

    Almost 50 papers presented were showing the status of the national programs related to conversion of research reactor cores from highly enriched (HEU) to low enriched uranium (LEU) fuel elements. Design of new fuel elements (uranium silicides) and safety related calculations were dealt with taking into account fuel cycle issues, meaning spent fuel storage and transportation. A number of presentations were devoted to Mo-99 production using LEU targets

  1. Virtual Averaging Making Nonframe-Averaged Optical Coherence Tomography Images Comparable to Frame-Averaged Images

    OpenAIRE

    Chen, Chieh-Li; Ishikawa, Hiroshi; Wollstein, Gadi; Bilonick, Richard A.; Kagemann, Larry; Schuman, Joel S.

    2016-01-01

    Purpose Developing a novel image enhancement method so that nonframe-averaged optical coherence tomography (OCT) images become comparable to active eye-tracking frame-averaged OCT images. Methods Twenty-one eyes of 21 healthy volunteers were scanned with noneye-tracking nonframe-averaged OCT device and active eye-tracking frame-averaged OCT device. Virtual averaging was applied to nonframe-averaged images with voxel resampling and adding amplitude deviation with 15-time repetitions. Signal-to...

  2. Uranium thorium dioxide fuel-cycle and economic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gale, J.D.; Spetz, S.W. [Framatome ANP, Inc., Lynchburg, Va. (United States)

    2001-07-01

    The fuel division of Framatome ANP (Advanced Nuclear Power) is performing a fuel-cycle analysis for uranium-thorium dioxide (U/Th) reactor fuel as part of a U.S. Department of Energy Nuclear Energy Research Initiative project titled, ''Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactor'', (DE-FC03-99SF21916). The objective is to evaluate the economic viability of the U/Th fuel cycle in commercial nuclear reactors operating in the U.S. This analysis includes formulating the evaluation methodology, validating the methodology via benchmark calculations, and performing a fuel-cycle analysis and corresponding economic evaluation. The APOLLO2-F computer program of Framatome ANP SCIENCE package was modified to incorporate the thorium decay chains and provide cross sections for the SCIENCE fuel-cycle analysis. A comparison and economic evaluation was made between UO{sub 2} and UO{sub 2}/ThO{sub 2} fuel cycles in a typical 193-fuel assembly pressurized water reactor using reload batch sizes corresponding to batch average discharge burnups of 50, 70, and 90 GWd/mtHM. Results show an increase in front-end costs for the UO{sub 2}/ThO{sub 2} cycles due primarily to the higher cost in separative work units for enriching the uranium to 19.5 wt% {sup 235}U. (author)

  3. HTGR fuel and fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740/sup 0/C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000/sup 0/C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-/sup 233/U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized.

  4. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U3O8-aluminum cermets. Above the aluminum melting point, U3O8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U3O8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 9000C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U3O8-aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 6600C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 9170C, while 7 psi average axial stress produced failure at 6690C

  5. TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel

    International Nuclear Information System (INIS)

    The experimental results of criticality benchmark experiments performed at the Jozef Stefan Institute TRIGA Mark II reactor are presented. The experiments were performed with partly burned fuel in two compact and uniform core configurations in the same arrangements as were used in the fresh fuel criticality benchmark experiment performed in 1991. In the experiments, both core configurations contained only 12 wt% U-ZrH fuel with 20% enriched uranium. The first experimental core contained 43 fuel elements with average burnup of 1.22 MWd or 2.8% 235U burned. The last experimental core configuration was composed of 48 fuel elements with average burnup of 1.15 MWd or 2.6% 235U burned. The experimental determination of keff for both core configurations, one subcritical and one critical, are presented. Burnup for all fuel elements was calculated in two-dimensional four-group diffusion approximation using the TRIGLAV code. The burnup of several fuel elements was measured also by the reactivity method

  6. Uranium enrichment. Principles

    International Nuclear Information System (INIS)

    Uranium enrichment industry is a more than 60 years old history and has developed without practically no cost, efficiency or profit constraints. However, remarkable improvements have been accomplished since the Second World War and have led to the development of various competing processes which reflect the diversity of uranium compositions and of uranium needs. Content: 1 - general considerations: uranium isotopes, problem of uranium enrichment, first realizations (USA, Russia, Europe, Asia, other countries), present day situation, future needs and market evolution; 2 - principles of isotopic separation: processes classification (high or low enrichment), low elementary enrichment processes, equilibrium time, cascade star-up and monitoring, multi-isotopes case, uranium reprocessing; 3 - enrichment and proliferation. (J.S.)

  7. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle (AECL-6594)

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium - 1.2 weight % U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed

  8. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium: 1.2 weight percent U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed. (author)

  9. Average Convexity in Communication Situations

    NARCIS (Netherlands)

    Slikker, M.

    1998-01-01

    In this paper we study inheritance properties of average convexity in communication situations. We show that the underlying graph ensures that the graphrestricted game originating from an average convex game is average convex if and only if every subgraph associated with a component of the underlyin

  10. The difference between alternative averages

    Directory of Open Access Journals (Sweden)

    James Vaupel

    2012-09-01

    Full Text Available BACKGROUND Demographers have long been interested in how compositional change, e.g., change in age structure, affects population averages. OBJECTIVE We want to deepen understanding of how compositional change affects population averages. RESULTS The difference between two averages of a variable, calculated using alternative weighting functions, equals the covariance between the variable and the ratio of the weighting functions, divided by the average of the ratio. We compare weighted and unweighted averages and also provide examples of use of the relationship in analyses of fertility and mortality. COMMENTS Other uses of covariances in formal demography are worth exploring.

  11. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Brim, Cornelia P.

    2013-04-01

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  12. Enriched uranium recovery at Los Alamos

    International Nuclear Information System (INIS)

    Graphite casting scrap, fuel elements and nongraphite combustibles are calcined to impure oxides. These materials along with zircaloy fuel elements and refractory solids are leach-dissolved separately in HF-HNO3 acid to solubilize the contained enriched uranium. The resulting slurry is filtered and the clear filtrate (to which mineral acid solutions bearing enriched uranium may be added) are passed through solvent extraction. The solvent extraction product is filtered, precipitated with H2O2 and the precipitate calcined to U3O8. Metal is made from U3O8 by conversion to UO2, hydrofluorination and reduction to metal. Throughput is 150 to 900 kg uranium per year depending on the type of scrap

  13. Status of RERTR fuel demonstrations

    International Nuclear Information System (INIS)

    A near-term objective of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is to demonstrate that the use of reduced-enrichment fuels meets the criteria of reliability, performance, safety, core lifetime, and economics. Two types of demonstrations are planned to meet this objective: fuel element irradiation testing and whole-core demonstrations. Data related to the first three criteria will come primarily from the element irradiations, whereas data related to the latter two, and, to some extent safety, will come from the whole-core demonstrations. The fuel element irradiations which discussed in this paper will be limited to those anticipated to be accomplished in the near term. The fuel types to be tested are UAlx-Al and U3O8-Al dispersions for plate-type reactors and U-ZrHx for rod-type reactors. The test fuel elements are being procured from CERCA (France), NUKEM (Germany), Texas Instruments (USA), and General Atomic Company (USA). It is planned that the irradiations will take place in the Oak Ridge Research Reactor (ORR), the High Flux Reactor at Petten (HFR-Petten, The Netherlands), the SILOE reactor (France), and the steady State Reactor (SSR, Romania). The latter is a new 14-MW TRIGA reactor. A tentative schedule for the irradiations and postirradiation examinations is shown. The burnups levels planned refer to average depletion of the 235-U originally contained in the fresh element. The goal of 75% burnup really represents achieving twice the fluence needed for 50% burnup. That level of burnup should certainly demonstrate that the reliability criterion has been achieved. Postirradiation examinations are planned for all of the types of plate-type elements. Visual inspections will be conducted in the reactor pool following irradiation. It is planned, for those elements irradiated in the ORR, to try to detect if any fission products are being released from the elements. After sufficient cooling time the elements will be transferred to a hot

  14. Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel

    CERN Document Server

    Hayes, A C

    2012-01-01

    The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

  15. 77 FR 823 - Guidance for Fuel Cycle Facility Change Processes

    Science.gov (United States)

    2012-01-06

    ... possess greater than a critical mass of special nuclear material and that are engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched uranium hexafluoride... Information DG-3037 was published in the Federal Register on July 14, 2011 (76 FR 41527). The public...

  16. The influence of fuel assembly characteristics on reactor safety

    International Nuclear Information System (INIS)

    To improve fuel utilization and nuclear plant economy, most nuclear plants of China adopt increased fuel enrichment and long cycle analysis. Core power distribution will be worse with these advanced items. Radial and axial peak increase too. This is a challenge to reactor safety. Since the fuel assembly is the most important part of a reactor core, fuel assembly characteristics affect reactor safety a lot. A few aspects of influence on reactor safety are discussed in this paper as a reference for fuel assembly design. A better fuel assembly design can increase heat exchange ability, especially in cold wall cells. The grids nearby core outlet can efficiently mix the flow of hot channel and average channel to decrease DNBR. In safety analysis, we always suppose the center of center assembly is the hot channel, but sometimes based on actual power distribution the hot channel occurs at side cell or corner cell. So the distribution of grids pressure drop coefficients can affect the minimum DNBR. A better fuel assembly design can help to spread core power distribution, decrease radial and axial peak efficiently. To spread core power distribution, different neutronic poisons are added into fuel pellet by different ways, and then the relative effects on reactor safety are different. At the same time, better fuel assembly design should leave enough margins for reactor safety to handle high burnup condition and so on. Fuel pellet and clad capabilities are getting worse versus increasing fuel burnup. This is a challenge to reactor safety, so more attentions should be paid to fuel burnup characteristics. (author)

  17. Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

    OpenAIRE

    Gruber, N.; Graven, H. D.

    2011-01-01

    Since aged carbon in fossil fuel contains no 14C, 14C/C ratios (Δ14C) measured in atmospheric CO2 can be used to estimate CO2 added by combustion and, potentially, provide verification of fossil CO2 emissions calculated using economic inventories. Sources of 14C from nuclear power generation and spent fuel reprocessing can counteract dilution by fossil CO2. Therefore, these nuclear sources can bias observation-based estimates of fossil fuel-derived CO2 if they are not correctly ...

  18. FUEL ASSAY REACTOR

    Science.gov (United States)

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  19. Data feature: Fuel procurement

    International Nuclear Information System (INIS)

    This document is a review of the effect of fuel costs on the procurement strategies of a utility and a conjecture that the same strategies may have an effect on the price of fuel. Factors affecting fuel costs are reviewed, and a number of procurement strategies taken to trim fuel costs are reviewed. The major trend is away from long-term enrichment contracts and into such strategies as: (1) Spot market purchases, (2) Inventory reduction, (3) Purchase of CIS material, and (4) Market-related contracts instead of base-escalated contracts

  20. Physical Theories with Average Symmetry

    CERN Document Server

    Alamino, Roberto C

    2013-01-01

    This Letter probes the existence of physical laws invariant only in average when subjected to some transformation. The concept of a symmetry transformation is broadened to include corruption by random noise and average symmetry is introduced by considering functions which are invariant only in average under these transformations. It is then shown that actions with average symmetry obey a modified version of Noether's Theorem with dissipative currents. The relation of this with possible violations of physical symmetries, as for instance Lorentz invariance in some quantum gravity theories, is briefly commented.

  1. Study of the U/sub 3/O/sub 8/-Al thermite reaction and strength of reactor fuel tubes

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H B

    1983-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U/sub 3/O/sub 8/-aluminum cermets. Above the aluminum melting point, U/sub 3/O/sub 8/ and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U/sub 3/O/sub 8/ in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900/sup 0/C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U/sub 3/O/sub 8/-aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660/sup 0/C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917/sup 0/C, while 7 psi average axial stress produced failure at 669/sup 0/C.

  2. How is uranium supply affecting enrichment?

    International Nuclear Information System (INIS)

    As a result of the enlivened uranium market, momentum has in turn picked up in the enrichment sector. What are the consequences of higher uranium prices? There is, of course, a link between uranium and enrichment supply to the extent that they are at least partial substitutes. On the enrichment supply side, the most obvious feature is the gradual replacement of the old gas diffusion facilities of Usec in the USA and EURODIF in France with more modern and economical centrifuge plants. Assuming Usec can overcome the financing and technical issues surrounding its plans, the last gas diffusion capacity should disappear around 2015 and the entire enrichment market should then be using centrifuges. On the commercial side, the key anticipated developments are mostly in Russia. Although there should still continue to be substantial quantities of surplus Russian HEU available for down blending in the period beyond 2013, it is now reasonable to expect that it will be mostly consumed by internal needs, to fuel Russian-origin reactors both at home and in export markets such as China and India. Finally, as a key sensitive area for the non-proliferation of nuclear weapons, the enrichment sector is likely to be a central point of the new international arrangements which must be developed to support a buoyant nuclear sector throughout this century.

  3. The physics scheme design for In-Hospital Neutron Irradiator-Mark 1 reactor loaded low-enrichment fuel%低浓化医院中子照射器(IHNI-1)堆芯的物理方案设计

    Institute of Scientific and Technical Information of China (English)

    江新标; 张文首; 高集金; 李义国; 周永茂

    2009-01-01

    采用蒙特卡罗程序MCNP/4B模拟计算了功率为30 kw的低浓化医院中子照射器的堆芯物理参数,设计了合理的堆芯布置方案、~(235)U富集度、控制棒价值、后备反应性和停堆深度,得到固有安全性较高、寿期达10年且无需换料、采用低浓化UO_2燃料的医院中子照射器的堆芯物理设计方案,为后续反应堆工程设计以及硼中子俘获治疗肿瘤用中子束的设计提供理论依据.%The core physics properties of 30 Kw In-Hospital Neutron Irradiator-Mark l(IHNI-l) reactor loaded low-enrichment fuel are simulated by using MCNP/4B Monte Carlo code. The arrangement scheme of core, ~(235)U enrichment, control rods' value, excess reactivity and shutdown margin are also reasonably introduced. The results show that the physics scheme of IHNI-1 reactor which possesses some particular characteristics, such as high inherent safety, 10 years lifetime without any refueling and low-enrichment UO_2 fuel, has been discovered in the paper and it will provide theoretic basis for reactor engineering design and neutron beam design for Boron Neutron Capture Therapy in the future.

  4. "Pricing Average Options on Commodities"

    OpenAIRE

    Kenichiro Shiraya; Akihiko Takahashi

    2010-01-01

    This paper proposes a new approximation formula for pricing average options on commodities under a stochastic volatility environment. In particular, it derives an option pricing formula under Heston and an extended lambda-SABR stochastic volatility models (which includes an extended SABR model as a special case). Moreover, numerical examples support the accuracy of the proposed average option pricing formula.

  5. Quantized average consensus with delay

    NARCIS (Netherlands)

    Jafarian, Matin; De Persis, Claudio

    2012-01-01

    Average consensus problem is a special case of cooperative control in which the agents of the network asymptotically converge to the average state (i.e., position) of the network by transferring information via a communication topology. One of the issues of the large scale networks is the cost of co

  6. Enrichment: Dealing with overcapacity

    International Nuclear Information System (INIS)

    Today's surplus of enrichment capacity will continue until at least the end of this century. This will challenge the ingenuity of the separative work unit (SWU) suppliers as they attempt to keep market share and remain profitable in a very competitive marketplace. The utilities will be faced with attractive choices, but making the best choice will require careful analysis and increased attention to market factors. Current demand projections will probably prove too high to the extent that more reactors are canceled or delayed. The DOE has the vast majority of the unused capacity, so it will feel the most immediate impact of this large surplus in productive capacity. The DOE has responded to these market challenges by planning another reorganization of its enriching operations. Without a major agreement among the governments affected by the current surplus in enrichment capacity, the future will see lower prices, more competitive terms, and the gradual substitution of centrifuge or laser enrichment for the gaseous diffusion plants. The competition that is forcing the gaseous diffusion prices down to marginal cost will provide the long-term price basis for the enrichment industry

  7. Gaussian moving averages and semimartingales

    DEFF Research Database (Denmark)

    Basse-O'Connor, Andreas

    2008-01-01

    In the present paper we study moving averages (also known as stochastic convolutions) driven by a Wiener process and with a deterministic kernel. Necessary and sufficient conditions on the kernel are provided for the moving average to be a semimartingale in its natural filtration. Our results...... are constructive - meaning that they provide a simple method to obtain kernels for which the moving average is a semimartingale or a Wiener process. Several examples are considered. In the last part of the paper we study general Gaussian processes with stationary increments. We provide necessary and sufficient...

  8. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  9. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm3 was by then in routine use, illustrated how far work has progressed

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To reconstruct a BWR type reactor into a high conversion reactor with no substantial changes for the reactor inner structure such as control rod structure. Constitution: The horizontal cross sectional shape of a channel box is reformed into a square configuration and the arrangement of fuel rods is formed as a trigonal lattice-like configuration. As a method of improving the conversion ratio, there is considered to use a dense lattice by narrowing the distance between fuel rods and trigonal lattice arrangement for fuel rod is advantageous therefor. A square shape cross sectional configuration having equal length both in the lateral and longitudinal directions is suitable for the channel box as a guide upon movement of the control rod. Fuel rods can be arranged with no loss by the trigonal lattice configuration, by which it is possible to improve the neutron moderation, increase the reactor core reactivity and conduct effective fuel combustion. In this way, it is possible to attain the object by inserting the follower portion of the control rod at the earier half and extracting the same at the latter half during the operation period in the reactor core comprising fuel assemblies suitable to a high conversion BWR type reactor having average conversion ratio of about 0.8. (Kamimura, M.)

  11. Laser and gas centrifuge enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Heinonen, Olli [Senior Fellow, Belfer Center for Science and International Affairs, Harvard Kennedy School, Cambridge, Massachusetts (United States)

    2014-05-09

    Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

  12. Laser and gas centrifuge enrichment

    Science.gov (United States)

    Heinonen, Olli

    2014-05-01

    Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

  13. Profile of World Uranium Enrichment Programs-2009

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    2009-04-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, whereas HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use as fuel for nuclear reactors to generate electricity. However, the same equipment used to produce LEU for nuclear reactor fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is not diverted or enriched to HEU. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 56 million kilogram separative work units (SWU) per year, with 22.5 million in gaseous diffusion and more than 33 million in gas centrifuge plants. Another 34 million SWU/year of capacity is under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future but has yet to be

  14. Segal Enriched Categories I

    CERN Document Server

    Bacard, Hugo V

    2010-01-01

    We develop a theory of enriched categories over a (higher) category M equipped with a class W of morphisms called homotopy equivalences. We call them Segal M_W -categories. Our motivation was to generalize the notion of "up-to-homotopy monoids" in a monoidal category M, introduced by Leinster. The formalism adopted generalizes the classical Segal categories and extends the theory of enriched category over a bicategory. In particular we have a linear version of Segal categories which did not exist so far. Our goal in this paper is to present the theory and provide some examples. Applications are reserved for the future.

  15. Advanced uranium enrichment processes

    International Nuclear Information System (INIS)

    Three advanced Uranium enrichment processes are dealt with in the report: AVLIS (Atomic Vapour LASER Isotope Separation), MLIS (Molecular LASER Isotope Separation) and PSP (Plasma Separation Process). The description of the physical and technical features of the processes constitutes a major part of the report. If further presents comparisons with existing industrially used enrichment technologies, gives information on actual development programmes and budgets and ends with a chapter on perspectives and conclusions. An extensive bibliography of the relevant open literature is added to the different subjects discussed. The report was drawn up by the nuclear research Centre (CEA) Saclay on behalf of the Commission of the European Communities

  16. Power convergence of Abel averages

    OpenAIRE

    Kozitsky, Yuri; Shoikhet, David; Zemanek, Jaroslav

    2012-01-01

    Necessary and sufficient conditions are presented for the Abel averages of discrete and strongly continuous semigroups, $T^k$ and $T_t$, to be power convergent in the operator norm in a complex Banach space. These results cover also the case where $T$ is unbounded and the corresponding Abel average is defined by means of the resolvent of $T$. They complement the classical results by Michael Lin establishing sufficient conditions for the corresponding convergence for a bounded $T$.

  17. Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Haitao, E-mail: haitaoju@gmail.com [Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan 610041 (China); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, The University of Tokyo, Hongo, Bunkyo, Tokyo 113-8656 (Japan); Oka, Yoshiaki [Joint Department of Nuclear Energy, Waseda University, Totsukamachi, Shinjuku, Tokyo 169-8050 (Japan)

    2015-08-15

    Highlights: • The improved core of Super Fast Reactor with high power density is analyzed. • We analyzed four types of the limiting fuel rods. • The influence of Pu enrichment and compressive stress to yield strength ratio are analyzed. • The improved fuel rod design of the new core is suggested. - Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should

  18. The effect of changing enrichments on core performance

    International Nuclear Information System (INIS)

    Highlights: • Five cores were analyzed with the same core configuration but with higher enrichments. • The highest enrichment core produced longest possible cycle length of 750 days. • New method for designing BP placement was introduced for the longer cycle lengths. • Fuel costs were calculated showing fuel costs decrease with increasing cycle length. - Abstract: The information presented in this paper has been developed as a follow on to two previous papers published using the same low leakage core configuration with the addition in this paper of evaluating fuel costs. The two previous publications studied the characteristics of this low leakage core with two different enrichment sets, where each enrichment set represents the three batches in the core. The purpose of the two previous papers proved the effectiveness of using the Haling Power Depletion (HPD) method as a guide. The first purpose of this paper is to extend this study to higher enrichment sets to finally attain a core having close to the highest possible cycle length. Three additional similar enrichment sets are studied increasing the number of enrichment sets to five. The ratio between the enrichment sets was maintained constant except for the highest enrichment set. This was done to increase the cycle length to approximately the longest possible cycle length of 800 days for a 1000 MWe reactor limited to a maximum 5% enrichment. The core reactor physics characteristics of these five cores are presented in this paper together with the evaluating of the fuel costs. These core characteristics include radial power fractions (RPF), Haling Power Depletion, RPF distributions, maximum pin peak powers (PPPMAX), and other important data. The HPD RPFs of all 5 cores were similar and used to help develop the burnable poison placement designs for each core. The longest two cycles required an improved technique using more information than the HPD results to develop successful BP placement designs. Also, it

  19. Enriching Number Knowledge

    Science.gov (United States)

    Mack, Nancy K.

    2011-01-01

    Exploring number systems of other cultures can be an enjoyable learning experience that enriches students' knowledge of numbers and number systems in important ways. It helps students deepen mental computation fluency, knowledge of place value, and equivalent representations for numbers. This article describes how the author designed her…

  20. Designing job enrichment projects.

    Science.gov (United States)

    Clakeley, G L

    1988-01-01

    This paper describes a management strategy for a job satisfaction program utilized in a large occupational therapy department. The goal of the program is to retain satisfied, productive employees and reduce attrition of therapists and assistants. The use of job enrichment projects for occupational therapy assistants will be presented with brief descriptions of two projects. PMID:23944880

  1. Job Enrichment in Extension.

    Science.gov (United States)

    Fourman, Louis S.; Jones, Jo

    1997-01-01

    Interviews with 10 participants in Ohio State University's job enrichment program for midcareer extension agents found that 5 returned to their same jobs after the experience but only 2 felt challenged/renewed. Part-time participation while working made it difficult to balance responsibilities. More information and a structured orientation were…

  2. Enriching the Catalog

    Science.gov (United States)

    Tennant, Roy

    2004-01-01

    After decades of costly and time-consuming effort, nearly all libraries have completed the retrospective conversion of their card catalogs to electronic form. However, bibliographic systems still are really not much more than card catalogs on wheels. Enriched content that Amazon.com takes for granted--such as digitized tables of contents, cover…

  3. Methodology for content enrichment

    NARCIS (Netherlands)

    Nederbragt, H.; Heerlien, M.

    2010-01-01

    The STERNA project mainly focuses on enrichment of existing content of content holding organisations in the natural history domain. Therefore, developing a methodology on how to best integrate one’s content into the STERNA information space is an essential part of the project. This document is the o

  4. Application of genetic algorithms and CASMO to fuel optimization of BWRs

    International Nuclear Information System (INIS)

    It was developed a system for the optimization of the radial distribution of enrichment in a fuel cell of a boiling water reactor based on genetic algorithms (GA's). The objective function includes four parameters: Average of the cell enrichment, average of gadolinium concentration of the cell, radial peak power factor and multiplication k-infinite factor. In order to be able to calculate the parameters that take part in the objective function, the process of evaluation of GA's was tied to the code CASMO-4, which is a code of transport in neutronic simulation groups of fuel assemblies that have been validated and it is used thoroughly for the calculation of nuclear data banks for boiling water reactors. A good radial distribution of fuel rods looks for, with different enrichment of U235 and contents of consumable poison in gadolinium form. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution problem. The optimization process was codified in language C in the operating system LINUX. It was automated the creation of the entrances to the simulator, the execution of simulator CASMO-4 and the obtaining of the parameters that take part in the objective function from the exit of the simulator. It was applied to the fuel cell design of lOxlO that can be used in the fuel designs which are used at the moment in the nuclear power plant of Laguna Verde. They were considered 10 different fuel compositions from which four contain gadolinium. Three heuristic rules were applied: the peripheral positions of the fuel cell cannot contain burning poison, are placed the compositions with the smallest enrichment in the cell corners and, it is fixed the placement of the water rods. Nevertheless, the placement of the rods with gadolinium cell inside left free. Designs were obtained that complete with the wanted reactivity and the radial peak power factor. The best in them has

  5. Production of Mo-99 using low-enriched uranium silicide

    International Nuclear Information System (INIS)

    Over the last several years, uranium silicide fuels have been under development as low-enriched uranium (LEU) targets for Mo-99. The use of LEU silicide is aimed at replacing the UAlx alloy in the highly-enriched uranium dissolution process. A process to recover Mo-99 from low-enriched uranium silicide is being developed at Argonne National Laboratory. The uranium silicide is dissolved in alkaline hydrogen peroxide. Experiments performed to determine the optimum dissolution procedure are discussed, and the results of dissolving a portion of a high-burnup (>40%) U3Si2 miniplate are presented. Future work related to Mo-99 separation and waste disposal are also discussed

  6. Validating MCNP for LEU Fuel Design via Power Distribution Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Maldonado, G Ivan [ORNL; Chandler, David [ORNL

    2008-11-01

    The mission of the Reduced Enrichment for Research and Test Reactors (RERTR) Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low enriched uranium (LEU) fuel and targets. Oak Ridge National Lab (ORNL) is reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction of flux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. A current 3-D Monte Carlo N-Particle (MCNP) model was modified to replicate the HFIR Critical Experiment 3 (HFIRCE-3) core of 1965. In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. Foils (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil s activity to the activity of a normalizing foil. The current work consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the normalizing foil. Power distributions were obtained for the clean core (no poison in moderator and symmetrical rod position at 17.5 inches) and fully poisoned-moderator (1.35 g B/liter in moderator and rods fully withdrawn) conditions. The observed deviations between the

  7. Standard specification for uranium metal enriched to more than 15 % and less Than 20 % 235U

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any 235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section 3. The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. In addition to this specification, the parties concerned may agree to other appropriate conditions. ...

  8. Comparisons of significant parameters for a standard 20% enriched and FLIP 70% enriched TRIGA core

    International Nuclear Information System (INIS)

    A comparison is made between the 20% and 70% enriched cores. The initial start-up data for both cores show the FLIP needs ∼3.8 times the 235U mass as the 20% core just to go critical. Operational configurations for both cores indicate a need for ∼33% additional fuel above initial critical for adequate maneuvering excess. The fuel element worths are higher in the central core locations for the 20% elements while the peripheral element worths are about the same (with some thermal flux peaking in the FLIP perheral elements). Pulsing comparisons of the two cores show significant differences in reactivity insertions and power peaks. (author)

  9. The Study of Numerical Simulation of Oxygen-‎enriched Burner System

    Directory of Open Access Journals (Sweden)

    Yuesheng Fan

    2010-12-01

    Full Text Available In order to reduce overall fuel consumption, or partially substitute a “valuable” fuel with a ‎poor one, in electric power plant boilers, oxygen enrichment of combustion air can be very ‎effective. The paper proposes an oxygen-enriched ignition system which based on the ‎existing pulverized coal fired boiler ignition devices. Small coal particle is suitable for this ‎system. The new burner includes inside, outside and middle casings. And it transfer heat in ‎two ways of downstream and upstream. The burner has authorized a patent in China. A ‎numerical simulation theory were used to analysis it. The results indicate that: it can ‎increase the maximum burning velocity ‎ ‎ and the average burning ‎velocity ‎, and decrease ignition temperature Ti and burnout temperature Tb of ‎pulverized coal. In addition, the pulverized coal fired boilers are easier to be ignited and the ‎comprehensive combustibility index S is improved. At the same time, it demonstrates that it ‎is an effective way to warm-up the pulverized coal in ignition of the boiler in the power ‎plant.‎

  10. Vocal attractiveness increases by averaging.

    Science.gov (United States)

    Bruckert, Laetitia; Bestelmeyer, Patricia; Latinus, Marianne; Rouger, Julien; Charest, Ian; Rousselet, Guillaume A; Kawahara, Hideki; Belin, Pascal

    2010-01-26

    Vocal attractiveness has a profound influence on listeners-a bias known as the "what sounds beautiful is good" vocal attractiveness stereotype [1]-with tangible impact on a voice owner's success at mating, job applications, and/or elections. The prevailing view holds that attractive voices are those that signal desirable attributes in a potential mate [2-4]-e.g., lower pitch in male voices. However, this account does not explain our preferences in more general social contexts in which voices of both genders are evaluated. Here we show that averaging voices via auditory morphing [5] results in more attractive voices, irrespective of the speaker's or listener's gender. Moreover, we show that this phenomenon is largely explained by two independent by-products of averaging: a smoother voice texture (reduced aperiodicities) and a greater similarity in pitch and timbre with the average of all voices (reduced "distance to mean"). These results provide the first evidence for a phenomenon of vocal attractiveness increases by averaging, analogous to a well-established effect of facial averaging [6, 7]. They highlight prototype-based coding [8] as a central feature of voice perception, emphasizing the similarity in the mechanisms of face and voice perception.

  11. Vocal attractiveness increases by averaging.

    Science.gov (United States)

    Bruckert, Laetitia; Bestelmeyer, Patricia; Latinus, Marianne; Rouger, Julien; Charest, Ian; Rousselet, Guillaume A; Kawahara, Hideki; Belin, Pascal

    2010-01-26

    Vocal attractiveness has a profound influence on listeners-a bias known as the "what sounds beautiful is good" vocal attractiveness stereotype [1]-with tangible impact on a voice owner's success at mating, job applications, and/or elections. The prevailing view holds that attractive voices are those that signal desirable attributes in a potential mate [2-4]-e.g., lower pitch in male voices. However, this account does not explain our preferences in more general social contexts in which voices of both genders are evaluated. Here we show that averaging voices via auditory morphing [5] results in more attractive voices, irrespective of the speaker's or listener's gender. Moreover, we show that this phenomenon is largely explained by two independent by-products of averaging: a smoother voice texture (reduced aperiodicities) and a greater similarity in pitch and timbre with the average of all voices (reduced "distance to mean"). These results provide the first evidence for a phenomenon of vocal attractiveness increases by averaging, analogous to a well-established effect of facial averaging [6, 7]. They highlight prototype-based coding [8] as a central feature of voice perception, emphasizing the similarity in the mechanisms of face and voice perception. PMID:20129047

  12. 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)

  13. Sparsity Averaging for Compressive Imaging

    CERN Document Server

    Carrillo, Rafael E; Van De Ville, Dimitri; Thiran, Jean-Philippe; Wiaux, Yves

    2012-01-01

    We propose a novel regularization method for sparse image reconstruction from compressive measurements. The approach relies on the conjecture that natural images exhibit strong average sparsity over multiple coherent frames. The associated reconstruction algorithm, based on an analysis prior and a reweighted $\\ell_1$ scheme, is dubbed Sparsity Averaging Reweighted Analysis (SARA). We test our prior and the associated algorithm through extensive numerical simulations for spread spectrum and Gaussian acquisition schemes suggested by the recent theory of compressed sensing with coherent and redundant dictionaries. Our results show that average sparsity outperforms state-of-the-art priors that promote sparsity in a single orthonormal basis or redundant frame, or that promote gradient sparsity. We also illustrate the performance of SARA in the context of Fourier imaging, for particular applications in astronomy and medicine.

  14. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  15. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  16. Uranium enrichment management review. Final report

    International Nuclear Information System (INIS)

    The uranium enrichment enterprise of the US Department of Energy (DOE) provides enriched nuclear fuel for private and government utilities domestically and abroad. The enterprise, in effect, provides a commercial service and represents a signficant business operation within the US government: more than $1 billion in revenues annually and future capital expenditures estimated at several billions of dollars. As a result, in May 1980, the Assistant Secretary for Resource Applications within DOE requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. The review group was specifically asked to focus on the management activities to which sound business practices could be applied. The group developed findings and recommendations in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. The chapters of this report present first the management review group's recommendations in the six areas evaluated and then the findings and issues in each area. An appendix provides the group's calendar of meetings. A list of major reference sources used in the course of the study is also included. 12 references

  17. Test bed for high-uranium-loaded fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1979-01-01

    An irradiation test facility has been designed and built to provide a test bed for irradiating a variety of miniature fuel plates. The objective of these tests is to screen various candidate materials as to their suitability for replacing the fully enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low-uranium enrichment (up to 45%) in place of highly-enriched fuel for these reactors would reduce the potential for /sup 235/U diversion.

  18. Postirradiation examination of high-U-loaded low-enriched U/sub 3/O/sub 8/, UAl/sub 2/, and U/sub 3/Si test fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

    1984-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U/sub 3/O/sub 8/, UAl/sub 2/ and U/sub 3/Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U/sub 3/Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup.

  19. Uranium Enrichment Reduction in the PGSFR Core Design

    International Nuclear Information System (INIS)

    Korea is currently developing the so-called Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) to investigate and demonstrate the capability of TRU transmutation. However, since fuel recycling technology is still at early development in Korea and also due to lack of experience in burning TRU in a fast reactor, the initial core of PGSFR is loaded with low-enriched uranium (LEU) fuel. Several test assemblies containing TRU fuels are supposed to be irradiated and tested for future TRU fuel developments. The uranium enrichment in the LEU PGSFR core is high, about 19.20%, due to large neutron leakage and low conversion ratio. In this paper, the required uranium enrichment is reduced by replacing the reflector material and modifying the reflector geometry in order to decrease the fuel cost of the LEU PGSFR core. PbO is chosen as the reflector material to replace the current HT9 and an inverted reflector assembly is also investigated in this study. It is shown that longer cycle length, higher fuel burnup and flattening power distribution can be achieved with PbO reflector and enhanced neutron leakage can be handled by the optimization of shielding material or core geometry. PbO reflector with inverted geometry is suggest in this research and by using inverted PbO reflector, core performance can be improved while leakage is negligibly enhanced than conventional pin type reflector assembly. Research about reducing the uranium enrichment more by increasing the uranium content in the uranium fuel which is U-10Zr now or increasing the smeared density which is currently 75% can be considered as a future work. Detailed analysis about multi-batch fuel management should be carried out since currently it is done approximately by using linear reactivity theory. Also, analysis for PGSFR with various reflector materials like LME, liquid lead will be carried out and the chemical reaction of those materials including PbO with sodium should be carefully investigated

  20. US developments in technology for uranium enrichment

    International Nuclear Information System (INIS)

    The purpose of this paper is to review recent progress and the status of the work in the United States on that part of the fuel cycle concerned with uranium enrichment. The United States has one enrichment process, gaseous diffusion, which has been continuously operated in large-scale production for the past 37 years; another process, gas centrifugation, which is now in the construction phase; and three new processes, molecular laser isotope separation, atomic vapor laser isotope separation, plasma separation process, in which the US has also invested sizable research and development efforts over the last few years. The emphasis in this paper is on the technical aspects of the various processes, but the important economic factors which will define the technological mix which may be applied in the next two decades are also discussed

  1. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

  2. Verification of 235U mass content in nuclear fuel plates by an absolute method

    Science.gov (United States)

    El-Gammal, W.

    2007-01-01

    Nuclear Safeguards is referred to a verification System by which a State can control all nuclear materials (NM) and nuclear activities under its authority. An effective and efficient Safeguards System must include a system of measurements with capabilities sufficient to verify such NM. Measurements of NM using absolute methods could eliminate the dependency on NM Standards, which are necessary for other relative or semi-absolute methods. In this work, an absolute method has been investigated to verify the 235U mass content in nuclear fuel plates of Material Testing Reactor (MTR) type. The most intense gamma-ray signature at 185.7 keV emitted after α-decay of the 235U nuclei was employed in the method. The measuring system (an HPGe-spectrometer) was mathematically calibrated for efficiency using the general Monte Carlo transport code MCNP-4B. The calibration results and the measured net count rate were used to estimate the 235U mass content in fuel plates at different detector-to-fuel plate distances. Two sets of fuel plates, containing natural and low enriched uranium, were measured at the Fuel Fabrication Facility. Average accuracies for the estimated 235U masses of about 2.62% and 0.3% are obtained for the fuel plates containing natural and low enriched uranium; respectively, with a precision of about 3%.

  3. Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

    Science.gov (United States)

    Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

    2016-10-01

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

  4. Data mining in the study of nuclear fuel cells

    International Nuclear Information System (INIS)

    In this paper is presented a study of data mining application in the analysis of fuel cells and their performance within a nuclear boiling water reactor. A decision tree was used to fulfill questions of the type If (condition) and Then (conclusion) to classify if the fuel cells will have good performance. The performance is measured by compliance or not of the cold shutdown margin, the rate of linear heat generation and the average heat generation in a plane of the reactor. It is assumed that the fuel cells are simulated in the reactor under a fuel reload and rod control patterns pre designed. 18125 fuel cells were simulated according to a steady-state calculation. The decision tree works on a target variable which is one of the three mentioned before. To analyze this objective, the decision tree works with a set of attribute variables. In this case, the attributes are characteristics of the cell as number of gadolinium rods, rods number with certain uranium enrichment mixed with a concentration of gadolinium, etc. The found model was able to predict the execution or not of the shutdown margin with a precision of around 95%. However, the other two variables showed lower percentages due to few learning cases of the model in which these variables were or were not achieved. Even with this inconvenience, the model is quite reliable and can be used in way coupled in optimization systems of fuel cells. (Author)

  5. Enrichment marketplace - today (and tomorrow)

    International Nuclear Information System (INIS)

    The technologies and capacities of the four primary sources of enrichment services, the United States Department of Energy, Eurodif, Techsnabexport of the Soviet Union, and Urenco, were given. Forecasts of future capacities and prices of enriched uranium were also included

  6. Dependability in Aggregation by Averaging

    CERN Document Server

    Jesus, Paulo; Almeida, Paulo Sérgio

    2010-01-01

    Aggregation is an important building block of modern distributed applications, allowing the determination of meaningful properties (e.g. network size, total storage capacity, average load, majorities, etc.) that are used to direct the execution of the system. However, the majority of the existing aggregation algorithms exhibit relevant dependability issues, when prospecting their use in real application environments. In this paper, we reveal some dependability issues of aggregation algorithms based on iterative averaging techniques, giving some directions to solve them. This class of algorithms is considered robust (when compared to common tree-based approaches), being independent from the used routing topology and providing an aggregation result at all nodes. However, their robustness is strongly challenged and their correctness often compromised, when changing the assumptions of their working environment to more realistic ones. The correctness of this class of algorithms relies on the maintenance of a funda...

  7. Stochastic Approximation with Averaging Innovation

    CERN Document Server

    Laruelle, Sophie

    2010-01-01

    The aim of the paper is to establish a convergence theorem for multi-dimensional stochastic approximation in a setting with innovations satisfying some averaging properties and to study some applications. The averaging assumptions allow us to unify the framework where the innovations are generated (to solve problems from Numerical Probability) and the one with exogenous innovations (market data, output of "device" $e.g.$ an Euler scheme) with stationary or ergodic properties. We propose several fields of applications with random innovations or quasi-random numbers. In particular we provide in both setting a rule to tune the step of the algorithm. At last we illustrate our results on five examples notably in Finance.

  8. High average power supercontinuum sources

    Indian Academy of Sciences (India)

    J C Travers

    2010-11-01

    The physical mechanisms and basic experimental techniques for the creation of high average spectral power supercontinuum sources is briefly reviewed. We focus on the use of high-power ytterbium-doped fibre lasers as pump sources, and the use of highly nonlinear photonic crystal fibres as the nonlinear medium. The most common experimental arrangements are described, including both continuous wave fibre laser systems with over 100 W pump power, and picosecond mode-locked, master oscillator power fibre amplifier systems, with over 10 kW peak pump power. These systems can produce broadband supercontinua with over 50 and 1 mW/nm average spectral power, respectively. Techniques for numerical modelling of the supercontinuum sources are presented and used to illustrate some supercontinuum dynamics. Some recent experimental results are presented.

  9. Fuel lattice design in a boiling water reactor using an ant-colony-based system

    International Nuclear Information System (INIS)

    Research highlights: → We present an ant-colony-based system for BWR fuel lattice design and optimization. → Assessment of candidate solutions at 0.0 MWd/kg 235U seems to have a limited scope. → Suitable heuristic rules enable more realistic fuel lattice designs. → The election of the objective has a large impact in CPU time. → ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U235 enrichment and Gd2O3 concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U235 enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U235 enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U235 enrichment; whereas, the k-infinity was inside the ±100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of

  10. Fuel lattice design in a boiling water reactor using an ant-colony-based system

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis, E-mail: joseluis.montes@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Facultad de Ciencias, Universidad Autonoma del Estado de Mexico (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Ortiz, Juan Jose, E-mail: juanjose.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Perusquia, Raul, E-mail: raul.perusquia@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico)

    2011-06-15

    Research highlights: > We present an ant-colony-based system for BWR fuel lattice design and optimization. > Assessment of candidate solutions at 0.0 MWd/kg {sup 235}U seems to have a limited scope. > Suitable heuristic rules enable more realistic fuel lattice designs. > The election of the objective has a large impact in CPU time. > ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U{sup 235} enrichment and Gd{sub 2}O{sub 3} concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U{sup 235} enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U{sup 235} enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U{sup 235} enrichment; whereas, the k-infinity was inside the {+-}100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a

  11. Surplus Highly Enriched Uranium Disposition Program plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    The purpose of this document is to provide upper level guidance for the program that will downblend surplus highly enriched uranium for use as commercial nuclear reactor fuel or low-level radioactive waste. The intent of this document is to outline the overall mission and program objectives. The document is also intended to provide a general basis for integration of disposition efforts among all applicable sites. This plan provides background information, establishes the scope of disposition activities, provides an approach to the mission and objectives, identifies programmatic assumptions, defines major roles, provides summary level schedules and milestones, and addresses budget requirements.

  12. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  13. FY2015 ceramic fuels development annual highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

    2015-09-22

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  14. The differential radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: accidental conditions

    International Nuclear Information System (INIS)

    The radiological impact of the fuel cycle of LWR type reactors using enriched uranium may be changed by plutonium recycle. The differences, which result from accidents which may occur in the different stages of the fuel cycle, are estimated in this study. The differential radiological impact on the population of the European Community is estimated for the recycle of 10t of plutonium metal, taking into consideration some characteristic accidents of each stage of the fuel cycle: fuel fabrication, reactor operation, fuel reprocessing and conversion, and, transport between the different units of the fuel cycle. Each unit is supposed built on an European ''average'' site (mean distributions of the populations and of the agricultural productions, reference meteorological situations). The recycle of plutonium in the fuel cycle involves a few per cent decrease of the radiological impact of the accident choosed for the nuclear power plants. The accidents of transport of plutonium, of new fuels and of plutonium wastes, as also thoses choosed for the fuel fabrication plant involve an increase of the impact for these types of transport and this plant. Finally, the differential radiological impact of the fuel reprocessing plant is positive but low

  15. Michel Parameters averages and interpretation

    International Nuclear Information System (INIS)

    The new measurements of Michel parameters in τ decays are combined to world averages. From these measurements model independent limits on non-standard model couplings are derived and interpretations in the framework of specific models are given. A lower limit of 2.5 tan β GeV on the mass of a charged Higgs boson in models with two Higgs doublets can be set and a 229 GeV limit on a right-handed W-boson in left-right symmetric models (95 % c.l.)

  16. Uranium enrichment by gas centrifuge

    International Nuclear Information System (INIS)

    After recalling the physical principles and the techniques of centrifuge enrichment the report describes the centrifuge enrichment programmes of the various countries concerned and compares this technology with other enrichment technologies like gaseous diffusion, laser, aerodynamic devices and chemical processes. The centrifuge enrichment process is said to be able to replace with advantage the existing enrichment facilities in the short and medium term. Future prospects of the process are also described, like recycled uranium enrichment and economic improvements; research and development needs to achieve the economic prospects are also indicated. Finally the report takes note of the positive aspect of centrifuge enrichment as far as safeguards and nuclear safety are concerned. 27 figs, 113 refs

  17. TMI Fuel Characteristics for Disposal Criticality Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Larry L. Taylor

    2003-09-01

    This report documents the reported contents of the Three Mile Island Unit 2 (TMI-2) canisters. proposed packaging, and degradation scenarios expected in the repository. Most fuels within the U.S. Department of Energy spent nuclear fuel inventory deal with highly enriched uranium, that in most cases require some form of neutronic poisoning inside the fuel canister. The TMI-2 fuel represents a departure from these fuel forms due to its lower enrichment (2.96% max.) values and the disrupted nature of the fuel itself. Criticality analysis of these fuel canisters has been performed over the years to reflect conditions expected during transit from the reactor to the Idaho National Engineering and Environmental Laboratory, water pool storage,1 and transport/dry-pack storage at Idaho Nuclear Technology and Engineering Center.2,3 None of these prior analyses reflect the potential disposal conditions for this fuel inside a postclosure repository.

  18. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  19. A Model to Predict Thermal Conductivity of Irradiated U-Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas; Huber, Tanja K.; Casella, Andrew M.

    2016-05-01

    The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

  20. A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

    Science.gov (United States)

    Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

    2016-05-01

    Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

  1. Advanced fuel developments to improve fuel cycle cost in PWR

    International Nuclear Information System (INIS)

    Increasingly lower fuel cycle costs and higher plant availability factors have been two crucial components in keeping the overall cost of electricity produced by nuclear low and competitive with respect to other energy sources. The continuous quest to reduce fuel cycle cost has resulted in some consolidated trends in LWR fuel management schemes: smaller number of feed fuel assemblies with longer residence time; longer cycles, with 18-month cycle as the predominant option, and some plants already operating on, or considering, 24-month refueling intervals; higher power ratings with many plants undergoing power uprates. In order to maintain or improve fuel utilization for the longer cycles and/or higher power ratings, the licensed limits in fuel fissile content (5.0 w/o U235 enrichment) and discharge burnup (62 GWd/tHM for the peak pin) have been approached. In addition, Zr-based fuel cladding materials are also being challenged by the resulting increased duty. For the above reasons further improvements in fuel cycle cost have to overcome one or more of the current limits. This paper discusses an option to break through this 'stalemate', i.e. uranium nitride (UN) fuel with SiC clad. In UN the higher density of the nitride with respect to the oxide fuel leads to higher fissile content and reduction in the number of feed assemblies, improved fuel utilization and potentially higher specific powers. The SiC clad, among other benefits, enables higher clad irradiation, thereby exploiting the full potential of UN fuel. An alternative to employing UN fuel is to maintain UO2 fuel but boost the fissile content increasing the U235 enrichment beyond the 5 w/o limit. The paper describes and compares the potential benefits on fuel cycle cost of either option using realistic full-core calculations and ensuing economic analysis performed using Westinghouse in-house reactor physics tools and methodologies. (author)

  2. Flexible time domain averaging technique

    Science.gov (United States)

    Zhao, Ming; Lin, Jing; Lei, Yaguo; Wang, Xiufeng

    2013-09-01

    Time domain averaging(TDA) is essentially a comb filter, it cannot extract the specified harmonics which may be caused by some faults, such as gear eccentric. Meanwhile, TDA always suffers from period cutting error(PCE) to different extent. Several improved TDA methods have been proposed, however they cannot completely eliminate the waveform reconstruction error caused by PCE. In order to overcome the shortcomings of conventional methods, a flexible time domain averaging(FTDA) technique is established, which adapts to the analyzed signal through adjusting each harmonic of the comb filter. In this technique, the explicit form of FTDA is first constructed by frequency domain sampling. Subsequently, chirp Z-transform(CZT) is employed in the algorithm of FTDA, which can improve the calculating efficiency significantly. Since the signal is reconstructed in the continuous time domain, there is no PCE in the FTDA. To validate the effectiveness of FTDA in the signal de-noising, interpolation and harmonic reconstruction, a simulated multi-components periodic signal that corrupted by noise is processed by FTDA. The simulation results show that the FTDA is capable of recovering the periodic components from the background noise effectively. Moreover, it can improve the signal-to-noise ratio by 7.9 dB compared with conventional ones. Experiments are also carried out on gearbox test rigs with chipped tooth and eccentricity gear, respectively. It is shown that the FTDA can identify the direction and severity of the eccentricity gear, and further enhances the amplitudes of impulses by 35%. The proposed technique not only solves the problem of PCE, but also provides a useful tool for the fault symptom extraction of rotating machinery.

  3. Burn-Up Calculations for the Brookhaven Graphite Research Reactor Fuel Elements

    International Nuclear Information System (INIS)

    Fuel bum-up calculations for the Brookhaven Graphite Research Reactor involve a distribution of the thermal megawatt days of operations to the fuel elements in proportion to the average thermal neutron flux at their location in the reactor. The megawatt days so assigned can be converted to equivalent uranium-235 consumption when needed. The original fuel loading for the BGRR was neutral uranium and a single calculation was performed on each fuel element upon discharge from the reactor. A subsequent change to a fully enriched uranium-235 fuel element, however, introduced complications. The average loading of enriched uranium involves about 4800 individual elements, each occupying four different reactor positions during its term in the reactor. The total term for a central channel element is about one year as against six to eight years for an element in a peripheral channel. With the large number of individual fuel elements involved and the approximately monthly small changes needed for operation, it was necessary to resort to a computer programme to follow the burn-up of all the elements on the reactor continuously. Both this and other functions of the computer programme are discussed in the paper. To date, uranium has been recovered from two batches of spent fuel. On the first, involving 3674 elements discharged from the reactor over a period of 4.9 years, the recovery figures were 5.5% higher than the calculated total of 32.3 kg uranium-235. On the second batch, involving 1296 elements discharged from the reactor over a period of one year, the recovery figures were 2.3% higher than the calculated figures of 10.8 kg uranium-235. This relatively close agreement seems to indicate that the assumptions made to simplify the programme are acceptable and that the results of the programme are satisfactory for our particular accounting and operating requirements. (author)

  4. DOE hands over uranium enrichment duties to government corporation

    International Nuclear Information System (INIS)

    In an effort to renew the United States' competitiveness in the world market for uranium enrichment services, the Department of Energy (DOE) is turning over control of its Paducah, KY, and Portsmouth, OH, enrichment facilities to a for-profit organization, the United States Enrichment Corp. (USEC), which was created by last year's Energy Policy Act. William H. Timbers, Jr., a former investment banker who was appointed acting CEO in March, said the Act's mandate will mean more competitive prices for enriched reactor fuel and greater responsiveness to utility customers. As a government corporation, USEC, with current annual revenues estimated at $1.5 billion, will no longer be part of the federal budget appropriations process, but will use business management techniques, set market-based prices for enriched uranium, and pay annual dividends to the US Treasury-its sole stockholder-from earnings. The goal is to finish privatizing the corporation within two years, and to sell its stock to investors for an estimated $1 to $3 billion. USEC's success will depend in part on developing short- and long-term marketing plants to help stanch the flow of enriched-uranium customers to foreign suppliers. (DOE already has received notice from a number of US utilities that they want to be let out of their long-term enrichment contracts as they expire over the next several years).USEC's plans likely will include exploring new joint ventures with other businesses in the nuclear fuel cycle-such as suppliers, fabricators, and converters-and offering a broader range of enrichment services than DOE provided. The corporation will have to be responsive to utilities on an individual basis

  5. Advanced Neutron Source enrichment study. Volume 2: Appendices -- Final report, Revision 12/94

    International Nuclear Information System (INIS)

    A study has been performed of the impact on performance of using low enriched uranium (20% 235U) or medium enriched uranium (35% 235U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% 235U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology. Volume 2 of this report contains 26 appendices containing results, meeting minutes, and fuel panel presentations. There are 26 appendices in this volume

  6. 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

  7. MOX Average Power Test 30 GWd/MT PIE: Quick Look

    Energy Technology Data Exchange (ETDEWEB)

    MORRIS, RN

    2001-02-14

    This report summarizes the early results of the post irradiation examination of the 30 GWd/MT MOX Average Power Test Capsules (numbers 3 and 10). The purpose of this preliminary examination is to document and monitor the progress of the MOX Average Power Test Irradiation. The capsules and their fuel pins were found to be in excellent condition. Measurement of the fission gas release fraction (about 1.50 to 2.26%), preliminary fuel stack gamma scan measurements, and preliminary fuel pin diameter measurements indicate that the fuel is behaving as expected.

  8. Automobile Fuel Economy: What is it Worth?

    OpenAIRE

    Nair, Santosh; Espey, Molly

    2004-01-01

    The marginal value of increased automobile fuel economy is estimated using a hedonic model of 2001 model year automobiles sold in the United States. This value is then compared to the average expected lifetime fuel savings attributable to increased fuel economy. Results indicate that automobile buyers fully internalize fuel cost savings attributable to improved fuel economy at low discount rates, and may partially internalize other perceived benefits of improved fuel economy such as reduction...

  9. Calculation study of TNPS spent fuel pool using burnup credit

    International Nuclear Information System (INIS)

    Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)

  10. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  11. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    Science.gov (United States)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

  12. U-Pu-Zr metallic fuel core and fuel concept for SFR with 550deg.C core outlet temperature

    International Nuclear Information System (INIS)

    The primary interest of SFR design core and fuel study in JAEA is to achieve 550 deg. C of core outlet temperature. The U-Pu-Zr metal fuel is an attractive fuel of SFR, which realizes the superior neutronic characteristic of the core due to its high content of heavy metal nuclide with appropriate irradiation experience of Zr alloy metal fuel. It is well understood that U-Pu-Zr base metal fuel has its major drawback as steel cladding temperature limiting feature due to fuel cladding compatibility concerns. Its cladding inner surface maximum temperature is to be limited as 650 deg. C to avoid the liquid phase formation in the fuel during steady state operation due to the inter-diffusion of elements (atoms) in the cladding and fuel. It is inevitable for the core thermal hydraulic design to give cladding maximum temperature in the core higher than core outlet temperature. A typical example of oxide fuel core design has 550 deg. C of core outlet temperature with 700 deg. C of cladding maximum temperature with uncertainties to be considered in the engineering design. Therefore, a core and fuel design with 550 deg. C of core outlet temperature and with 650 deg. C of cladding maximum temperature is significant challenge of SFR metallic fuel core study. Such core has been designed by significantly reducing the metallic fuel pin power variation during the irradiation. The core concept has single Pu enrichment which achieves local neutronic conversion ratio as close as 1.0, which leads to stable fuel pin power distribution in the core and stable fuel pin power history of each fuel pin. An example of 1,785MWt core which corresponds to 750MWe is as follows: Fuel pin diameter: 7.5 to 8.5 mm; Core fuel column length: 900 tp 1000mm; Pu enrichment: 12 %; Zr content 6 to 10wt.%; Fuel smeared density: 70 to 75%TD; Reactor operation cycle length: 24-26 months; Burnup reactivity swing: 0.5 % del k/kk'; Breeding ratio: 1.0 (without blanket); Fuel average burnup: 90 GWd/t. The results of

  13. Grade point average and biographical data in personal resumes: predictors of finding employment

    OpenAIRE

    Sulastri, A.; Handoko, M.; Janssens, J.M.A.M.

    2015-01-01

    This study aimed to examine relationships between graduates' grade point average (GPA), biographical data and success in finding a job in general and a psychology-based job in particular. Two hundred six psychology graduates participated in a two-wave longitudinal study. Biographical data assessed were extracurricular activities, computer and foreign language skills, participation in a growth mindset enrichment programme and in general enrichment programmes, and work experiences. GPA showed s...

  14. Feasible advanced fuel cycle options for CANDU reactors in the Republic of Korea

    International Nuclear Information System (INIS)

    Taking into account the view points on nuclear safety, nuclear waste, non-proliferation and economics from the public, international environment, and utilities, the SEU/RU and DUPIC fuel cycles would be feasible options of advanced fuel cycles for CANDU-PHWRs in the Republic of Korea in the mid- and long-terms, respectively. Comparing with NU fuel, 0.9 % or 1.2 % SEU fuel would increase fuel burnup and hence reduce the spent fuel arisings by a factor of 2 or 3, and also could reduce CANDU fuel cycle costs by 20 to 30%. RU offers similar benefits as 0.9% SEU and is very attractive due to the significantly improved fuel cycle economics, substantially increased burnups, large reduction in fuel requirements as well as in spent fuel arisings. For RU use in a CANDU reactor, re-enrichment is not required. There are 25,000 tes RU produced from reprocessing operations in Europe and Japan, which would theoretically provide sufficient fuel for 500 CANDU 6 reactor-years of operation. According to the physics, thermal-hydraulic and thermal-mechanical assessments of CANFLEX-0.9% RU fuel for a CANDU-6 reactor, the fuel could be introduced into the reactor in a straight-forward fashion. A series of assessments of CANFLEX-DUPIC physics on the compatibility of the fuel design in the existing CANDU 6 reactors has shown that the poisoning of the central element of DUPIC with, for example, natural dysprosium, reduces the void reactivity of the fuel, and that a 2 bundle shift refuelling scheme would be the most appropriate in-core fuel management scheme for a CANDU-6 reactor. The average discharge burnup is ∼15 MWd/kgHE. Although these results have shown promising results for the DUPIC fuel cycle, more in-depth studies are required in the areas of ROP system, large LOCA safety analyses, and so on. The recycling fuel cycles of RU and DUPIC for CANDU are expected to achieve the environmental 3R's (Reduce, Reuse, Recycle) as applied to global energy use in the short- and long

  15. Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database

  16. 78 FR 77650 - Low Enriched Uranium From France: Continuation of Antidumping Duty Order

    Science.gov (United States)

    2013-12-24

    ... defined as enriched uranium dioxide (UO 2 ), whether or not contained in nuclear fuel rods or assemblies... form, such as UO 2 , or fabricated into nuclear fuel assemblies, regardless of the means by which the... of the LEU for consumption by the end- user in a nuclear reactor outside the United States....

  17. 77 FR 19642 - Low Enriched Uranium From France: Final Results of Antidumping Duty Changed Circumstances Review

    Science.gov (United States)

    2012-04-02

    ... defined as enriched uranium dioxide (UO 2 ), whether or not contained in nuclear fuel rods or assemblies... of the LEU for consumption by the end- user in a nuclear reactor outside the United States. Such... form, such as UO 2 , or fabricated into nuclear fuel assemblies, regardless of the means by which...

  18. Earthquake forecast enrichment scores

    Directory of Open Access Journals (Sweden)

    Christine Smyth

    2012-03-01

    Full Text Available The Collaboratory for the Study of Earthquake Predictability (CSEP is a global project aimed at testing earthquake forecast models in a fair environment. Various metrics are currently used to evaluate the submitted forecasts. However, the CSEP still lacks easily understandable metrics with which to rank the universal performance of the forecast models. In this research, we modify a well-known and respected metric from another statistical field, bioinformatics, to make it suitable for evaluating earthquake forecasts, such as those submitted to the CSEP initiative. The metric, originally called a gene-set enrichment score, is based on a Kolmogorov-Smirnov statistic. Our modified metric assesses if, over a certain time period, the forecast values at locations where earthquakes have occurred are significantly increased compared to the values for all locations where earthquakes did not occur. Permutation testing allows for a significance value to be placed upon the score. Unlike the metrics currently employed by the CSEP, the score places no assumption on the distribution of earthquake occurrence nor requires an arbitrary reference forecast. In this research, we apply the modified metric to simulated data and real forecast data to show it is a powerful and robust technique, capable of ranking competing earthquake forecasts.

  19. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    The papers presented at the International Conference on The Nuclear Fuel Cycle, held at Stockholm, 28 to 31 October 1975, are reviewed. The meeting, organised by the U.S. Atomic Industrial Forum, and the Swedish Nuclear Forum, was concerned more particularly with economic, political, social and commercial aspects than with tecnology. The papers discussed were considered under the subject heading of current status, uranium resources, enrichment, and reprocessing. (U.K.)

  20. Progress in Japanese uranium enrichment technology and the promotion of industrialization

    International Nuclear Information System (INIS)

    The development of uranium enrichment technology with centrifuges, undertaken as a national project, has reached the stage that the pilot plant built by the Power Reactor and Nuclear Fuel Development Corporation will start the partial operation around the middle of this year. In order to assure the constant supply of nuclear fuel to Japan, it is necessary to establish an independent nuclear fuel cycle, and for the purpose, uranium enrichment must be industrialized by the use of Japanese own technology, because for the time being, light water reactors will be mainly employed. The Uranium Enrichment Evaluation Subcommittee was set up in July, 1978, and began the studies by evaluating the progress made in the development of centrifugal enrichment techniques by the PNC. The goal of development is changing from the technical improvement of centrifuges themselves to the economic performance. The facilities for the conversion of feed uranium into UF6, the blending and storage of enriched uranium, and the enrichment proper including centrifuge production were studied, and the existing fundamental technologies in Japan can solve the problems without much difficulty. The results of evaluation of the enrichment technology, the urgent need for uranium enrichment in Japan, and the measures to promote the industrialization are explained. (Kako, I.)

  1. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately

  2. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  3. Report of Sectional Committee on Industrialization of Uranium Enrichment

    International Nuclear Information System (INIS)

    In order to accelerate the development and utilization of atomic energy which is the core of the substitute energies for petroleum, it is indispensable requirement to establish independent fuel cycle as the base. In particular, the domestic production of enriched uranium is necessary to eliminate the obstacles to secure the energy supply in Japan. The construction and operation of the pilot plant for uranium enrichment by centrifugal separation method have progressed smoothly, and the technical base for the domestic production of enriched uranium is being consolidated. For the time being, the service of uranium enrichment is given by USA and France, but it is expected that the short supply will arise around 1990. The start of operation of the uranium enrichment plant in Japan is scheduled around 1990, and the scale of the plant will be expanded stepwise thereafter. The scale of production is assumed as 3000 t SWU/year in 2000. Prior to this commercial plant, the prototype plant of up to 250 t SWU/year capacity will be operated in 1986, starting the production of centrifugal separators in 1983. The production line for centrifugal separators will have the capacity of up to 125 t SWU/year. The organization for operating these plants, the home production of natural uranium conversion, the uranium enrichment by chemical method and others are described. (Kako, I.)

  4. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  5. Proceedings of the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program

  6. Thorium fuel cycle studies: fuel fabrication process and cost estimation

    International Nuclear Information System (INIS)

    Early in 1976 a study was made to assess the relative economics and fuel utilization of thorium and uranium fuel cycles in various types of reactors. It was to be completed in approximately two months, so all component parts had to be developed in a short time with a high degree of dependence on existing information. One of the components required for the study was a consistent set of relatively accurate fuel fabrication costs for the various reactor-fuel combinations. A report documents the rationale used in generating these cost estimates and presents in some detail the basis and methodology employed. Since three types of thermal flux reactors (LWR, HWR, and HTGR) and two types of fast flux reactors (liquid metal and gas cooled) together with three fuel forms (oxides, carbides, and metal) were included in the study with various combinations of the fissionable metals U, Th, and Pu, it was necessary to define a methodology that would permit a rapid relative estimate for each case. Existing cost studies were chosen for a Light-Water Reactor with low-enriched uranium fuel and for a High-Temperature Gas-Cooled Reactor with highly enriched uranium and thorium fuel as the reference cases which could be compared with other reactor-fuel combinations

  7. Fuel element development

    International Nuclear Information System (INIS)

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

  8. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    International Nuclear Information System (INIS)

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

  9. Fuel cycle of the AVR

    International Nuclear Information System (INIS)

    All the stages of development were secured by irradiation tests and by the use of the elements concerned in the AVR. Summarising, these were: The first charge with fuel elements from Union Carbide, with U-Th mixed carbide particles, wallpaper variants, U-Th mixed carbide particles, the pressed fuel element with U-Th mixed carbide particles, the pressed fuel element with U-Th mixed oxide particles, with an intermediate boundary layer; the present THTR element, the pressed fuel element with U-Th mixed oxide particles and low temperature PyC coating, the pressed fuel element with U-Th mixed oxide particles, with PyC and SiC layers, i.e.: TRISO particles, the pressed fuel element with pure uranium oxide particles for the low enrichment cycle, one coated only by 2 PyC layers, the other coated with PyC and SiC, i.e.: TRISO coating. (orig.)

  10. Future fuel cycles

    International Nuclear Information System (INIS)

    A fuel cycle must offer both financial and resource savings if it is to be considered for introduction into Ontario's nuclear system. The most promising alternative CANDU fuel cycles are examined in the context of both of these factors over a wide range of installed capacity growth rates and economic assumptions, in order to determine which fuel cycle, or cycles, should be introduced, and when. It is concluded that the optimum path for the long term begins with the prompt introduction of the low-enriched-uranium fuel cycle. For a wide range of conditions, this cycle remains the optimum throughout the very long term. Conditions of rapid nuclear growth and very high uranium price escalation rates warrant the supersedure of the low-enriched-uranium cycle by either a plutonium-topped thorium cycle or plutonium recycle, beginning between 2010 and 2025. It is also found that the uranium resource position is sound in terms of both known resources and production capability. Moreover, introduction of the low-enriched-uranium fuel cycle and 1250 MWe reactor units will assure the economic viability of nuclear power until at least 2020, even if uranium prices increase at a rate of 3.5% above inflation. The interrelationship between these two conclusions lies in the tremendous incentive for exploration which will occur if the real uranium price escalation rate is high. From a competitive viewpoint, nuclear power can withstand increases in the price of uranium. However, such increases will likely further expand the resource base, making nuclear an even more reliable energy source. (auth)

  11. Interpreting Sky-Averaged 21-cm Measurements

    Science.gov (United States)

    Mirocha, Jordan

    2015-01-01

    Within the first ~billion years after the Big Bang, the intergalactic medium (IGM) underwent a remarkable transformation, from a uniform sea of cold neutral hydrogen gas to a fully ionized, metal-enriched plasma. Three milestones during this epoch of reionization -- the emergence of the first stars, black holes (BHs), and full-fledged galaxies -- are expected to manifest themselves as extrema in sky-averaged ("global") measurements of the redshifted 21-cm background. However, interpreting these measurements will be complicated by the presence of strong foregrounds and non-trivialities in the radiative transfer (RT) modeling required to make robust predictions.I have developed numerical models that efficiently solve the frequency-dependent radiative transfer equation, which has led to two advances in studies of the global 21-cm signal. First, frequency-dependent solutions facilitate studies of how the global 21-cm signal may be used to constrain the detailed spectral properties of the first stars, BHs, and galaxies, rather than just the timing of their formation. And second, the speed of these calculations allows one to search vast expanses of a currently unconstrained parameter space, while simultaneously characterizing the degeneracies between parameters of interest. I find principally that (1) physical properties of the IGM, such as its temperature and ionization state, can be constrained robustly from observations of the global 21-cm signal without invoking models for the astrophysical sources themselves, (2) translating IGM properties to galaxy properties is challenging, in large part due to frequency-dependent effects. For instance, evolution in the characteristic spectrum of accreting BHs can modify the 21-cm absorption signal at levels accessible to first generation instruments, but could easily be confused with evolution in the X-ray luminosity star-formation rate relation. Finally, (3) the independent constraints most likely to aide in the interpretation

  12. 7 CFR 1209.12 - On average.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false On average. 1209.12 Section 1209.12 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS....12 On average. On average means a rolling average of production or imports during the last two...

  13. Application of genetic algorithms and CASMO to fuel optimization of BWRs; Aplicacion de algoritmos geneticos y CASMO a la optimizacion de combustible de BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Carmona H, R.; Martin del Campo M, C.; Oropeza C, I.P. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: rockbert@ieee.org

    2008-07-01

    It was developed a system for the optimization of the radial distribution of enrichment in a fuel cell of a boiling water reactor based on genetic algorithms (GA's). The objective function includes four parameters: Average of the cell enrichment, average of gadolinium concentration of the cell, radial peak power factor and multiplication k-infinite factor. In order to be able to calculate the parameters that take part in the objective function, the process of evaluation of GA's was tied to the code CASMO-4, which is a code of transport in neutronic simulation groups of fuel assemblies that have been validated and it is used thoroughly for the calculation of nuclear data banks for boiling water reactors. A good radial distribution of fuel rods looks for, with different enrichment of U{sup 2}35 and contents of consumable poison in gadolinium form. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution problem. The optimization process was codified in language C in the operating system LINUX. It was automated the creation of the entrances to the simulator, the execution of simulator CASMO-4 and the obtaining of the parameters that take part in the objective function from the exit of the simulator. It was applied to the fuel cell design of lOxlO that can be used in the fuel designs which are used at the moment in the nuclear power plant of Laguna Verde. They were considered 10 different fuel compositions from which four contain gadolinium. Three heuristic rules were applied: the peripheral positions of the fuel cell cannot contain burning poison, are placed the compositions with the smallest enrichment in the cell corners and, it is fixed the placement of the water rods. Nevertheless, the placement of the rods with gadolinium cell inside left free. Designs were obtained that complete with the wanted reactivity and the radial peak power factor. The

  14. Economic analysis of fuel recycle

    International Nuclear Information System (INIS)

    Economic analysis was performed at KAERI with the assistance of US DOE to compare single reactor fuel cycle costs for a once-through option and a thermal recycle option to operate 1 GWe of a PWR plant for its lifetime. A reference fuel cycle cost was first calculated for each option with best estimated reference input data. Then a sensitivity analysis was performed changing each single value of such fuel cycle component costs as yellow cake price, enrichment charges, spent fuel storage cost, reprocessing cost, spent fuel disposal cost and reprocessing waste disposal cost. Savings due to thermal recycle in requirements of uranium, conversion, and enrichment were examined using formulas suggested by US DOE, while MOX fabrication penalty was accounted for. As a result of the reference fuel cycle cost analysis, it is calculated that the thermal recycle option is marginally more economical than the once-through option. The major factors affecting the comparative costs between thermal recycle and once-through are the costs of reprocessing, spent fuel storage and the difference between spent fuel disposal and reprocessing waste disposal. However, considering the uncertainty in these cost parameters there seems no immediate economic incentive for thermal recycle at the present time

  15. Review of non-proliferable fuel options in research reactors (TRIGA)

    International Nuclear Information System (INIS)

    This study is to examine the following aspects resulted from uranium enrichment reduction in TRIGA Mark-III reactor: reactor performance, fuel cycle costs, safety and reliability, and non-proliferation aspects. Fuel models adopted are: existing fuels (20% and 70% enriched with 8.5 wt% U-loaded); proposed fuels (20% and 30% enriched with 12 wt% U-loaded, 40% enriched with 8.5 wt% U-loaded, and 20% enriched with 20 wt% U-loaded). As results, the proposed fuels are disadvantageous over current FLIP-fuelled core in safety margins, reliability and reactor performance. Besides, operating costs will double with 12 wt% U-loaded fuels than with FLIP fuels

  16. Current Status of the MLIS Uranium Enrichment Process

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Woo; Rhee, Chang Kyu; Kim, Whung Whoe [Nuclear Materials Research Division, Daejeon (Korea, Republic of); Yang, Maeng Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Eerkens, Jeff W. [Isotope Technologies Inc., Woodland (United States)

    2009-05-15

    Well-established energy security is extremely important for the national economy and future prosperity of the country. Many countries are trying to develop and use renewable energy sources such as solar, wind, and tidal power to keep the natural environment clean and safe. Although some have disputed calling nuclear power (which produces {approx}40% of domestic electricity) 'green energy', it deserves credit for this label because of its low CO{sub 2} emission and fuel efficiency. While fuel costs in nuclear power generation are less than 30% and uranium enrichment is only {approx}40% of front-end fuel loop, uranium enrichment is a critical step in nuclear power generation. It is also politically sensitive worldwide due to potential proliferation aspects. Currently, gaseous diffusion and centrifuge are the technologies for uranium enrichment. Since the 70s, many countries have been trying to develop a more advanced and economic technology after the gaseous diffusion process started to loose its economic viability. Among the developed laser-assisted technologies, MLIS (Molecular Laser Isotope Separation) is still under investigation for industrial application while most other laser uranium enrichment technologies have been terminated. In this regard it is worthwhile to review the MLIS processes being developed currently, SILEX (Separation of Isotopes by Laser EXcitation) and CRISLA (Condensation Repression by Isotope Selective Laser Activation). Recently the Canadian company Cameco joined the GE-Hitachi venture by paying $123.8M for a 24% holding, and GE-Hitachi- Cameco is now running the GLE (Global Laser Enrichment) facility using SILEX technology in Wilmington, North Carolina to evaluate commercialization of the process.

  17. Thorium fuel cycle study for PWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae Yong; Kim, Myung Hyun [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-12-31

    A nuclear design feasibility of thorium fueled high converting PWR was investigated. Two kinds of fuel design option were tested for the comparison with conventional UO{sub 2} fuel. The first one was an application of MHTGR pyro-carbon coated particle fuels. The other design was an application of MOX fuels as a ThO{sub 2}-PuO{sub 2} ceramic pellet. In the case of carbon-coated particle fuels, there was no benefit in nuclear design aspect because enrichment of U-235 was required over 5 w/o in order to match with the K-infinite of Ulchin-3/4 fuels. However, the use of thorium based plutonium fuels in PWR gave favorable aspects in nuclear design such as flatter K-infinite curve, lower M. T. C. and lower F. T. C. than that of UO{sub 2} fuel. (author). 6 refs., 3 tabs., 6 figs.

  18. Initial report on characterization of excess highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    DOE`s Office of Fissile Materials Disposition assigned to this Y-12 division the task of preparing a report on the 174.4 metric tons of excess highly enriched U. Characterization included identification by category, gathering existing data (assay), defining the likely needed processing steps for prepping for transfer to a blending site, and developing a range of preliminary cost estimates for those steps. Focus is on making commercial reactor fuel as a final disposition path.

  19. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  20. Economic study of fuel scenarios for a reload; Estudio economico de escenarios de combustible para una recarga

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, J. A.; Montes T, J. L.; Perusquia del C, R., E-mail: juanjose.ortiz@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this work the results to plan different scenarios for designing a nuclear fuel reload are shown. Given a reload with specific energy requirements, the objective is to verify the feasibility of using either a greater number of fresh fuel with less uranium enrichment, or otherwise reduce the number of fresh fuel assemblies and therefore they have a higher average uranium enrichment. For the study a cycle balance 18-month basis with 112 fresh assemblies divided into two lots, with energy produced of 10,075 Mwd/Tu was used. For the designs under the mentioned scenarios, the heuristic techniques known as taboo search and neural networks were used. To verify the feasibility of obtained reloads an economic study of the reload costs was performed. The results showed that is possible to design reloads under the two scenarios, but was more complicated decrease the amount of fresh fuel assemblies. In both scenarios was possible to reduce manufacturing costs of fuel and according to purely static calculation, it would be possible to increase the energy produced. (Author)

  1. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  2. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  3. Fast Reactor Fuel Development in Japan

    International Nuclear Information System (INIS)

    The future fast reactor and its fuel cycle system under development in Japan uses oxide fuel with simplified pelletizing fuel fabrication technology as a reference concept. Its driver fuel consists of large diameter annular fuel pellets, oxide dispersion strengthened ferritic steel cladding fuel pins with a ferritic-martensitic steel subassembly wrapper tube and minoractinide- bearing oxide fuel. The target burnup of the driver fuel is 150 GW.d/t in discharge average, which corresponds to 250 GW.d/t of peak burnup and 250 dpa of peak neutron dose. Fuel developmental efforts, including out-of-pile studies such as material characteristics experimental evaluation and fuel property measurements, various irradiation tests and fuel fabrication technology developments were planned and are in progress. Future fuels will be realized through Joyo irradiation tests and Monju demonstrations. International collaborative efforts are also an important part of such activities. (author)

  4. Meadow enriched ACP process algebras

    OpenAIRE

    J.A. Bergstra; Middelburg, C.A.

    2009-01-01

    We introduce the notion of an ACP process algebra. The models of the axiom system ACP are the origin of this notion. ACP process algebras have to do with processes in which no data are involved. We also introduce the notion of a meadow enriched ACP process algebra, which is a simple generalization of the notion of an ACP process algebra to processes in which data are involved. In meadow enriched ACP process algebras, the mathematical structure for data is a meadow.

  5. Standard specification for uranium hexafluoride enriched to less than 5 % 235U

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification covers nuclear grade uranium hexafluoride (UF6) that either has been processed through an enrichment plant, or has been produced by the blending of Highly Enriched Uranium with other uranium to obtain uranium of any 235U concentration below 5 % and that is intended for fuel fabrication. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Enriched Commercial Grade UF6 so that, with respect to fuel design and manufacture, it is essentially equivalent to enriched uranium made from natural UF6; and (2) To define limits for Enriched Reprocessed UF6 to be expected if Reprocessed UF6 is to be enriched without dilution with Commercial Natural UF6. For such UF6, special provisions, not defined herein, may be needed to ensure fuel performance and to protect the work force, process equipment, and the environment. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched UF6 that is to be used in the pro...

  6. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    Energy Technology Data Exchange (ETDEWEB)

    CHASTAIN, S.A.

    2005-10-24

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The

  7. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle is substantially more complicated than the energy production cycles of conventional fuels because of the very low abundance of uranium 235, the presence of radioactivity, the potential for producing fissile nuclides from irradiation, and the risk that fissile materials will be used for nuclear weapons. These factors add enrichment, recycling, spent fuel storage, and safeguards to the cycle, besides making the conventional steps of exploration, mining, processing, use, waste disposal, and transportation more difficult

  8. Irradiation testing of miniature fuel plates for the RERTR program

    International Nuclear Information System (INIS)

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (mini-plates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly-enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low-uranium enrichment of about 20% 235U in place of highly enriched fuel for these reactors would reduce the potential for 235U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminium-base dispersion-type fuel plates with fuel cores of: (1) high uranium content U3O8-Al being developed by ORNL; (2) high uranium content UAl/sub x/-Al being developed by EG and G Idaho, Inc.; and (3) very high uranium content U3Si-Al being developed by ANL. The irradiation test facility, designated as HFED-1, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. Ultimately, fuel plate types with suitable characteristics will be manufactured into full-sized plate-type fuel elements suitable for testing in the ORR. Specifications for these elements are described in Appendix A

  9. A premature demise for RERTR [Reduced Enrichment for Research and Test Reactors programme]?

    International Nuclear Information System (INIS)

    A common commitment from France, Belgium, Germany and the US to eliminate highly enriched uranium from their research reactors is needed to help guard against this material falling into the wrong hands. In the US, an essential part of this commitment would be rekindling the weakened Reduced Enrichment for Research and Test Reactors programme (RERTR). This is an American initiative to develop low-enrichment uranium fuel for research reactors that have previously required weapons-usable material. Underway since 1978 at Argonne National Laboratory, RERTR has achieved some impressive results: the development of higher density, low enriched fuels that are suitable for use at over 90% of the world's research reactors; a net reduction of US exports of highly enriched uranium (HEU) from the annual 700kg levels in the late 1970s to a 1990 level of just over 100kg; the encouragement of international scientific co-operation aimed at developing new fuels and facilitating the conversion of existing reactors to these fuels. However, in recent years, the US commitment to RERTR has been declining -budgets have fallen and advanced fuel development work has terminated. (author)

  10. Effectiveness of bioremediation for the Prestige fuel spill : a summary of case studies

    Energy Technology Data Exchange (ETDEWEB)

    Gallego, J.R. [Oviedo Univ., Asturias (Spain); Gonzalez-Rojas, E.; Pelaez, A.I.; Sanchez, J [Oviedo Univ., Asturias (Spain). Inst. de Biotecnologia de Asturias; Garcia-Martinez, M.J.; Llamas, J.F. [Univ. Polictenica de Madrid, Madrid (Spain). Laboratorio de Estratigrafia Biomolecular

    2006-07-01

    This paper described novel bioremediation strategies used to remediate coastal areas in Spain impacted by the Prestige fuel oil spill in 2002. The bioremediation techniques were applied after hot pressurized water washing was used to remove hydrocarbons adhering to shorelines and rocks. Bioremediation strategies included monitored natural attenuation as well as accelerating biodegradation by stimulating indigenous populations through the addition of exogenous microbial populations. The sites selected for bioremediation were rocky shorelines of heterogenous granitic sediments with grain sizes ranging from sands to huge boulders; limestone-sandstone pebbles and cobbles; and fuel-coated limestone cliffs. Total surface area covered by the fuel was determined through the use of image analysis calculations. A statistical measurement of the fuel layer thickness was calculated by averaging the weights of multiple-fuel sampling increments. Bioremediation products included the use of oleophilic fertilizers; a biodegradable surfactant; and a microbial seeding agent. Determinations of saturate, aromatic, resins, and asphaltene (SARA) were performed using maltenes extraction and liquid chromatography. Microbial plating and selective enrichment with fuel as the sole carbon source were used to monitor the evolution of microbial populations in a variety of experiments. It was concluded that the biostimulation technique enhanced the efficiency of the in situ oleophilic fertilizers. 17 refs., 2 tabs., 6 figs.

  11. Effectiveness of bioremediation for the Prestige fuel spill : a summary of case studies

    International Nuclear Information System (INIS)

    This paper described novel bioremediation strategies used to remediate coastal areas in Spain impacted by the Prestige fuel oil spill in 2002. The bioremediation techniques were applied after hot pressurized water washing was used to remove hydrocarbons adhering to shorelines and rocks. Bioremediation strategies included monitored natural attenuation as well as accelerating biodegradation by stimulating indigenous populations through the addition of exogenous microbial populations. The sites selected for bioremediation were rocky shorelines of heterogenous granitic sediments with grain sizes ranging from sands to huge boulders; limestone-sandstone pebbles and cobbles; and fuel-coated limestone cliffs. Total surface area covered by the fuel was determined through the use of image analysis calculations. A statistical measurement of the fuel layer thickness was calculated by averaging the weights of multiple-fuel sampling increments. Bioremediation products included the use of oleophilic fertilizers; a biodegradable surfactant; and a microbial seeding agent. Determinations of saturate, aromatic, resins, and asphaltene (SARA) were performed using maltenes extraction and liquid chromatography. Microbial plating and selective enrichment with fuel as the sole carbon source were used to monitor the evolution of microbial populations in a variety of experiments. It was concluded that the biostimulation technique enhanced the efficiency of the in situ oleophilic fertilizers. 17 refs., 2 tabs., 6 figs

  12. Averages of Values of L-Series

    OpenAIRE

    Alkan, Emre; Ono, Ken

    2013-01-01

    We obtain an exact formula for the average of values of L-series over two independent odd characters. The average of any positive moment of values at s = 1 is then expressed in terms of finite cotangent sums subject to congruence conditions. As consequences, bounds on such cotangent sums, limit points for the average of first moment of L-series at s = 1 and the average size of positive moments of character sums related to the class number are deduced.

  13. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    International Nuclear Information System (INIS)

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers

  14. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.

  15. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  16. Enrichment of desert soil elements in Takla Makan dust aerosol

    International Nuclear Information System (INIS)

    During a Hungarian expedition in 1994 to arid regions of north-western China, atmospheric aerosol samples were collected in the Takla Makan Desert and on some sites in mountains surrounding the Tarim Basin. PIXE data obtained for the composition and enrichment factors of the regional aerosol clearly reflected that a heavy accumulation of salts has been formed in the closed inland basin. When compared to the regional soil composition data published by other authors, it turned out that S and Cl, showing high enrichment relative to average crust composition, are of soil origin

  17. Spectral averaging techniques for Jacobi matrices

    CERN Document Server

    del Rio, Rafael; Schulz-Baldes, Hermann

    2008-01-01

    Spectral averaging techniques for one-dimensional discrete Schroedinger operators are revisited and extended. In particular, simultaneous averaging over several parameters is discussed. Special focus is put on proving lower bounds on the density of the averaged spectral measures. These Wegner type estimates are used to analyze stability properties for the spectral types of Jacobi matrices under local perturbations.

  18. Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

    International Nuclear Information System (INIS)

    Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

  19. Novel Membranes and Processes for Oxygen Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Haiqing

    2011-11-15

    The overall goal of this project is to develop a membrane process that produces air containing 25-35% oxygen, at a cost of $25-40/ton of equivalent pure oxygen (EPO2). Oxygen-enriched air at such a low cost will allow existing air-fueled furnaces to be converted economically to oxygen-enriched furnaces, which in turn will improve the economic and energy efficiency of combustion processes significantly, and reduce the cost of CO{sub 2} capture and sequestration from flue gases throughout the U.S. manufacturing industries. During the 12-month Concept Definition project: We identified a series of perfluoropolymers (PFPs) with promising oxygen/nitrogen separation properties, which were successfully made into thin film composite membranes. The membranes showed oxygen permeance as high as 1,200 gpu and oxygen/nitrogen selectivity of 3.0, and the permeance and selectivity were stable over the time period tested (60 days). We successfully scaled up the production of high-flux PFP-based membranes, using MTR's commercial coaters. Two bench-scale spiral-wound modules with countercurrent designs were made and parametric tests were performed to understand the effect of feed flow rate and pressure, permeate pressure and sweep flow rate on the membrane module separation properties. At various operating conditions that modeled potential industrial operating conditions, the module separation properties were similar to the pure-gas separation properties in the membrane stamps. We also identified and synthesized new polymers [including polymers of intrinsic microporosity (PIMs) and polyimides] with higher oxygen/nitrogen selectivity (3.5-5.0) than the PFPs, and made these polymers into thin film composite membranes. However, these membranes were susceptible to severe aging; pure-gas permeance decreased nearly six-fold within two weeks, making them impractical for industrial applications of oxygen enrichment. We tested the effect of oxygen-enriched air on NO{sub x} emissions

  20. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel (235U enrichment > 90%) with low-enriched uranium (LEU) fuel (235U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  1. Proliferation resistance fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Ko, W. I

    1999-02-01

    The issues of dual use in nuclear technology are analysed for nuclear fuel cycle with special focus on uranium enrichment and spent fuel reprocessing which are considered as the most sensitive components in terms of vulnerability to diversion. Technical alternatives to mitigrate the vulnerability, as has been analysed in depth during the NASAP and INFCE era in the late seventies, are reviewed to characterize the DUPIC fuel cycle alternative. On the other hand, the new realities in nuclear energy including the disposition of weapon materials as a legacy of cold war are recast in an angle of nuclear proliferation resistance and safeguards with a discussion on the concept of spent fuel standard concept and its compliance with the DUPIC fuel cycle technology. (author)

  2. Proliferation resistance fuel cycle technology

    International Nuclear Information System (INIS)

    The issues of dual use in nuclear technology are analysed for nuclear fuel cycle with special focus on uranium enrichment and spent fuel reprocessing which are considered as the most sensitive components in terms of vulnerability to diversion. Technical alternatives to mitigrate the vulnerability, as has been analysed in depth during the NASAP and INFCE era in the late seventies, are reviewed to characterize the DUPIC fuel cycle alternative. On the other hand, the new realities in nuclear energy including the disposition of weapon materials as a legacy of cold war are recast in an angle of nuclear proliferation resistance and safeguards with a discussion on the concept of spent fuel standard concept and its compliance with the DUPIC fuel cycle technology. (author)

  3. Full conversion of materials and nuclear fuel research and test - TRIGA SSR 14 MW

    International Nuclear Information System (INIS)

    This article presents the HEU (high enrichment uranium) to LEU (low enrichment uranium) conversion of the TRIGA reactor at the Institute for Nuclear Research (Pitesti, Romania). This process began in 1992 when the first 4 LEU (23% U235 enrichment) fuel bundles integrated the reactor core in replacement of 4 HEU fuel bundles. By March 2004, the mixed reactor core had 18 LEU and 17 HEU fuel bundles by HEU-LEU replacement through successive steps of fueling. In 2006 the conversion process was completed and now we have a standard reactor core of 29 LEU fuel bundles

  4. Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki

    1996-11-01

    In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)

  5. The Status and Development Potential of Plate-Type Fuels for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1979-03-01

    Recent U.S. Department of State action to restrict the shipment and use of highly enriched uranium for research and test reactors has renewed fuel development activity. The objective of these development activities is to increase the total uranium loading in the fuel meat so that enrichment reduction can be accomplished without significant performance penalties. This report characterizes the status and the potential for development of the currently utilized plate-type fuels for research and test reactors. The report also characterizes the newer high-density fuels which could be utilized in these reactors and indicates the impact of the utilization of both the new and current fuels on enrichment reduction.

  6. Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

    Energy Technology Data Exchange (ETDEWEB)

    Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge; Glenn A. Moore; Mitchell K. Meyer

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Mo fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the

  7. Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

    Energy Technology Data Exchange (ETDEWEB)

    Jue, Jan-Fong, E-mail: dennis.keiser@inl.gov; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U–Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U–10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: • A typical Zr diffusion barrier with a thickness of 25 μm. • A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 μm. • Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt.%. • Decomposed areas containing plate-shaped low-Mo phase. • A typical Zr/cladding interaction layer with a thickness of 1–2 μm. • A visible UZr{sub 2} bearing layer with a thickness of 1–2 μm. • Mo-rich precipitates (mainly Mo{sub 2}Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr{sub 2}-bearing layer and the U–Mo matrix. • No excessive interaction between cladding and the uncoated fuel edge. • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O

  8. Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

    Science.gov (United States)

    Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: diffusion barrier with a thickness of 25 μm. A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 μm. Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7-13 wt.%. Decomposed areas containing plate-shaped low-Mo phase. A typical Zr/cladding interaction layer with a thickness of 1-2 μm. A visible UZr2 bearing layer with a thickness of 1-2 μm. Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U-Mo matrix. No excessive interaction between cladding and the uncoated fuel edge. Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. Some of these attributes might be critical to the

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    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.

  10. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  11. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Dayman, Kenneth J., E-mail: kenneth.dayman@gmail.com [University of Texas at Austin, Austin, TX 78758 (United States); Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M. [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States)

    2014-01-21

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial {sup 235}U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra. -- Highlights: • We investigate characterization of used nuclear fuel with multivariate analysis. • PWR and BWR fuels expected in

  12. Enrichment of light hydrocarbon mixture

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Dali (Los Alamos, NM); Devlin, David (Santa Fe, NM); Barbero, Robert S. (Santa Cruz, NM); Carrera, Martin E. (Naperville, IL); Colling, Craig W. (Warrenville, IL)

    2011-11-29

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  13. A Systematic Approach to Marital Enrichment.

    Science.gov (United States)

    Dinkmeyer, Don; Carlson, Jon

    1986-01-01

    Presents a systematic approach to enriching marital relationships. The history and current status of marital enrichment is reviewed. An Adlerian approach to marital enrichment is described. Applications of the program in enrichment groups, marriage therapy and couple groups are included. (Author)

  14. Metal enrichment in a semi-analytical model, fundamental scaling relations, and the case of Milky Way galaxies

    Science.gov (United States)

    Cousin, M.; Buat, V.; Boissier, S.; Bethermin, M.; Roehlly, Y.; Génois, M.

    2016-05-01

    Context. Gas flows play a fundamental role in galaxy formation and evolution, providing the fuel for the star formation process. These mechanisms leave an imprint in the amount of heavy elements that enrich the interstellar medium. Thus, the analysis of this metallicity signature provides additional constraint on the galaxy formation scenario. Aims: We aim to discriminate between four different galaxy formation models based on two accretion scenarios and two different star formation recipes. We address the impact of a bimodal accretion scenario and a strongly regulated star formation recipe on the metal enrichment process of galaxies. Methods: We present a new extension of the eGalICS model, which allows us to track the metal enrichment process in both stellar populations and in the gas phase. Based on stellar metallicity bins from 0 to 2.5 Z⊙, our new chemodynamical model is applicable for situations ranging from metal-free primordial accretion to very enriched interstellar gas contents. We use this new tool to predict the metallicity evolution of both the stellar populations and gas phase. We compare these predictions with recent observational measurements. We also address the evolution of the gas metallicity with the star formation rate (SFR). We then focus on a sub-sample of Milky Way-like galaxies. We compare both the cosmic stellar mass assembly and the metal enrichment process of such galaxies with observations and detailed chemical evolution models. Results: Our models, based on a strong star formation regulation, allow us to reproduce well the stellar mass to gas-phase metallicity relation observed in the local Universe. The shape of our average stellar mass to stellar metallicity relations is in good agreement with observations. However, we observe a systematic shift towards high masses. Our M⋆ - Zg -SFR relation is in good agreement with recent measurements: our best model predicts a clear dependence with the SFR. Both SFR and metal enrichment

  15. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  16. Uranium enrichment. 1980 annual report

    International Nuclear Information System (INIS)

    This report contains data and related information on the production of enriched uranium at the gaseous diffusion plants and an update on the construction and project control center for the gas centrifuge plant. Power usage at the gaseous diffusion plants is illustrated. The report contains several glossy color pictures of the plants and processes described. In addition to gaseous diffusion and the centrifuge process, three advanced isotope separation process are now being developed. The business operation of the enrichment plants is described; charts on revenue, balance sheets, and income statements are included

  17. Thermogenic Effect from Nutritionally Enriched Coffee Consumption

    Directory of Open Access Journals (Sweden)

    Jennings Peter F

    2006-06-01

    Full Text Available Abstract Background The purpose of this study was to examine the effect of nutritionally enriched JavaFit™ (JF coffee (450 mg of caffeine, 1200 mg of garcinia cambogia, 360 mg of citrus aurantium extract, and 225 mcg of chromium polynicotinate on resting oxygen uptake (VO2, respiratory exchange ratio (RER, heart rate (HR, and blood pressure (BP in healthy and physically active individuals. Method Ten subjects (8 male, 2 female; 20.9 ± 1.7 y; 178.1 ± 10.4 cm; 71.8 ± 12.1 kg underwent two testing sessions administered in a randomized and double-blind fashion. During each session, subjects reported to the Human Performance Laboratory after at least 3-h post-absorptive state and were provided either 354 ml (1.5 cups of freshly brewed JF or commercially available caffeinated coffee (P. Subjects then rested in a semi-recumbent position for three hours. VO2 and HR were determined every 5 min during the first 30 min and every 10 min during the next 150 min. BP was determined every 15 min during the first 30 min and every 30 min thereafter. Area under the curve (AUC analysis was computed for VO2, whereas a session-average was calculated for RER, HR and BP. Results Initial analysis revealed no significant differences. However, seven of the ten subjects were considered responders to JF (had a higher AUC for VO2during JF than P. Statistical analysis showed the difference between JF and P (12% to be significantly different in these responders. In addition, the average systolic BP was higher (p Conclusion It appears that consuming a nutritionally-enriched coffee beverage may increase resting energy expenditure in individuals that are sensitive to the caffeine and herbal coffee supplement. In addition, this supplement also appears to affect cardiovascular dynamics by augmenting systolic arterial blood pressure.

  18. Average-cost based robust structural control

    Science.gov (United States)

    Hagood, Nesbitt W.

    1993-01-01

    A method is presented for the synthesis of robust controllers for linear time invariant structural systems with parameterized uncertainty. The method involves minimizing quantities related to the quadratic cost (H2-norm) averaged over a set of systems described by real parameters such as natural frequencies and modal residues. Bounded average cost is shown to imply stability over the set of systems. Approximations for the exact average are derived and proposed as cost functionals. The properties of these approximate average cost functionals are established. The exact average and approximate average cost functionals are used to derive dynamic controllers which can provide stability robustness. The robustness properties of these controllers are demonstrated in illustrative numerical examples and tested in a simple SISO experiment on the MIT multi-point alignment testbed.

  19. MEASUREMENT AND MODELLING AVERAGE PHOTOSYNTHESIS OF MAIZE

    OpenAIRE

    ZS LÕKE

    2005-01-01

    The photosynthesis of fully developed maize was investigated in the Agrometeorological Research Station Keszthely, in 2000. We used LI-6400 type measurement equipment to locate measurement points where the intensity of photosynthesis mostly nears the average. So later we could obtain average photosynthetic activities featuring the crop, with only one measurement. To check average photosynthesis of maize we used Goudriaan’s simulation model (CMSM) as well to calculate values on cloudless sampl...

  20. WIDTHS AND AVERAGE WIDTHS OF SOBOLEV CLASSES

    Institute of Scientific and Technical Information of China (English)

    刘永平; 许贵桥

    2003-01-01

    This paper concerns the problem of the Kolmogorov n-width, the linear n-width, the Gel'fand n-width and the Bernstein n-width of Sobolev classes of the periodicmultivariate functions in the space Lp(Td) and the average Bernstein σ-width, averageKolmogorov σ-widths, the average linear σ-widths of Sobolev classes of the multivariatequantities.

  1. NOAA Average Annual Salinity (3-Zone)

    Data.gov (United States)

    California Department of Resources — The 3-Zone Average Annual Salinity Digital Geography is a digital spatial framework developed using geographic information system (GIS) technology. These salinity...

  2. Stochastic averaging of quasi-Hamiltonian systems

    Institute of Scientific and Technical Information of China (English)

    朱位秋

    1996-01-01

    A stochastic averaging method is proposed for quasi-Hamiltonian systems (Hamiltonian systems with light dampings subject to weakly stochastic excitations). Various versions of the method, depending on whether the associated Hamiltonian systems are integrable or nonintegrable, resonant or nonresonant, are discussed. It is pointed out that the standard stochastic averaging method and the stochastic averaging method of energy envelope are special cases of the stochastic averaging method of quasi-Hamiltonian systems and that the results obtained by this method for several examples prove its effectiveness.

  3. The Study of Numerical Simulation of Oxygen-‎enriched Burner System

    OpenAIRE

    Yuesheng Fan; Pengfei Si

    2010-01-01

    In order to reduce overall fuel consumption, or partially substitute a “valuable” fuel with a ‎poor one, in electric power plant boilers, oxygen enrichment of combustion air can be very ‎effective. The paper proposes an oxygen-enriched ignition system which based on the ‎existing pulverized coal fired boiler ignition devices. Small coal particle is suitable for this ‎system. The new burner includes inside, outside and middle casings. And it transfer heat in ‎two ways of downstream and upstrea...

  4. Meadow enriched ACP process algebras

    NARCIS (Netherlands)

    J.A. Bergstra; C.A. Middelburg

    2009-01-01

    We introduce the notion of an ACP process algebra. The models of the axiom system ACP are the origin of this notion. ACP process algebras have to do with processes in which no data are involved. We also introduce the notion of a meadow enriched ACP process algebra, which is a simple generalization o

  5. Environmental Development Plan: uranium enrichment

    International Nuclear Information System (INIS)

    This Environmental Development Plan identifies and examines the environmental, health, safety, and socioeconomic concerns and corresponding requirements associated with the DOE research, development, demonstration, and operation of the Uranium Enrichment program, including the gaseous diffusion process, the centrifuge process, centrifuge rotor fabrication, and related research and development activities

  6. Evaluation of irradiation swelling behaviour of RSG-GAS fuel

    International Nuclear Information System (INIS)

    Evaluation of irradiation swelling behaviour of RSG-GAS fuel has been performed. The evaluation is intended to support the fuel utilization improvement program by increasing the fuel discharge burnup as high as possible. Fuel that has been evaluated is dispersion type of low enriched uranium of U3Si2-Al (2.96 gr U/cm3) in the form of straight plate (MTR box shape fuel element). The evaluation is based on DART calculation results for steady-state irradiation at RSG-GAS nominal power (30 MW thermal). Up to fuel burnup of 100% U235, the effect of fuel swelling for steady state irradiation on the mechanical integrity of fuel cladding and the thermal performance of the fuel can be ignored (within the design tolerance). This indicates that the silicide fuel being used in RSG-GAS basically can be irradiated until fuel burnup as high as 100% U235. (author)

  7. A working plan for working group 2 'enrichment' within the scope of INFCE

    International Nuclear Information System (INIS)

    A working plan for INFCE/WG.2 is presented, outlining the major questions which the group needs to answer under the headings: 1. Enrichment needs and supply, 2. Models for cross-investment, 3. Market situation, 4. Technical and economic assessment of the different enrichment technologies, and 5. Safeguards aspects. It is suggested that the group's assessment should include: 1. Future enrichment capacities, 2. Multinational or regional fuel cycle centres, 3. Possible patterns for guarantees of supply, and 4. Special needs of developing countries

  8. Safety of Conversion Facilities and Uranium Enrichment Facilities. Specific Safety Guide

    International Nuclear Information System (INIS)

    This Safety Guide supplements the Safety Requirements publication on Safety of Fuel Cycle Facilities and addresses all the stages in the life cycle of conversion facilities (CFs) and enrichment facilities (EFs), with emphasis placed on design and operation. It describes the actions, conditions and procedures for meeting safety requirements and deals specifically with the handling, processing and storage of depleted, natural and low enriched uranium. The publication is intended to be of use to designers, operating organizations and regulators for ensuring the safety of conversion and enrichment facilities. Contents: 1. Introduction; 2. General safety recommendations; 3. Site evaluation; 4. Design; 5. Construction; 6. Commissioning; 7. Operation; 8. Decommissioning; Annexes.

  9. Average Transmission Probability of a Random Stack

    Science.gov (United States)

    Lu, Yin; Miniatura, Christian; Englert, Berthold-Georg

    2010-01-01

    The transmission through a stack of identical slabs that are separated by gaps with random widths is usually treated by calculating the average of the logarithm of the transmission probability. We show how to calculate the average of the transmission probability itself with the aid of a recurrence relation and derive analytical upper and lower…

  10. Average sampling theorems for shift invariant subspaces

    Institute of Scientific and Technical Information of China (English)

    孙文昌; 周性伟

    2000-01-01

    The sampling theorem is one of the most powerful results in signal analysis. In this paper, we study the average sampling on shift invariant subspaces, e.g. wavelet subspaces. We show that if a subspace satisfies certain conditions, then every function in the subspace is uniquely determined and can be reconstructed by its local averages near certain sampling points. Examples are given.

  11. Average excitation potentials of air and aluminium

    NARCIS (Netherlands)

    Bogaardt, M.; Koudijs, B.

    1951-01-01

    By means of a graphical method the average excitation potential I may be derived from experimental data. Average values for Iair and IAl have been obtained. It is shown that in representing range/energy relations by means of Bethe's well known formula, I has to be taken as a continuously changing fu

  12. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    fuel produce the same amount of energy as the conventional uranium-based fuel. The results of the studies show that in the ThPu and in the MOX fuels partially significant smaller quantities of most of the long-lived actinides are produced. The time-averaged power profile along the active height in the assemblies loaded with the heterogeneous fuel is broader distributed and the maximum temperature in both fuel types is approximately 250 K lower than the maximum temperature in the UO2 fuel. This yields to a decrease in the Doppler broadening of the resonance lines in the absorption cross section as well as to smaller temperature gradients in the material structures of the assembly. The fuel burnup is correlated with the power profile, hence, an axially more constant fuel utilization in comparison to uranium-based fuels with a homogeneous enrichment is observed. Regarding the initial activity of the spent fuels, it is shown that the contribution of the fission products to the total activity is negligible in the case of ThPu and MOX which is contrary to the UO2 fuel, where the total activity is dominated by the fission products at the very beginning of the storage time. However, due to the high fraction of plutonium in the innovative fuels, the total activity decreases slower than the total activity of the UO2 fuel by a factor of two (MOX) and 2.5 (ThPu). On the other hand, this behavior is typical for conventional mixed oxide fuels. At core level, the uranium-based fuel assemblies are replaced with 11% and 18% of the assemblies heterogeneously loaded with the MOX and the ThPu fuels, respectively. An optimum loading pattern is achieved with a minimum peaking factor for both core models. The assemblies loaded with the innovative fuels influence the residual UO2 fuel assemblies and the operational behavior of the entire core by (amongst others) an extension of the operational cycle of about 22%. This extension is explained as follows: for the utilization of - MOX: an

  13. Bacterioplankton responses to iron enrichment during the SAGE experiment

    Science.gov (United States)

    Kuparinen, J.; Hall, J.; Ellwood, M.; Safi, K.; Peloquin, J.; Katz, D.

    2011-03-01

    We studied the microbial food web in the upper 100 m of the water column in iron-limited sub-Antarctic HNLC waters south-east of New Zealand in the SAGE experiment in 2004, with focus on bacterioplankton. Samples were collected daily from inside and outside the iron enriched patch. Short term enrichment experiments were conducted on board in 4 L polycarbonate bottles with water outside the iron enriched patch to study single and combined effects of micronutrient additions on microbial food web. Low bacterial growth was recorded in the study area with community turnover times of 50 h or more during the study period. Measurements of bacterial standing stocks and production rates in the study show minor responses to the large scale iron enrichment, with increase in rates and stocks after the first enrichment and at the end of the study period after the third iron enrichment when solar radiation increased and wind mixing decreased. The average daily bacterial production rates were 31.5 and 33.7 mgCm -2 d -1 for the OUT and IN stations, respectively; thus overall there was not a significant difference between the control and the iron-enriched patch. In the bottle experiments bacterial thymidine incorporation showed responses to single iron and silicic acid enrichments and a major growth response to the combined iron and sucrose enrichments. Phytoplankton chlorophyll- a showed clear stimulation by single additions of iron and silicic acid and silicic acid enhanced the iron impact. Cobalt additions had no effect on bacteria growth and a negative effect on phytoplankton growth. Low bacterial in situ growth rates and the enrichment experiments suggest that bacteria are co-limited by iron and carbon, and that bacterial iron uptake is dependent on carbon supply by the food web. With the high iron quota (μmol Fe mol C -1) bacteria may scavenge considerable amounts of the excess iron, and thus influence the relative importance of the microbial food web as a carbon sink.

  14. New results on averaging theory and applications

    Science.gov (United States)

    Cândido, Murilo R.; Llibre, Jaume

    2016-08-01

    The usual averaging theory reduces the computation of some periodic solutions of a system of ordinary differential equations, to find the simple zeros of an associated averaged function. When one of these zeros is not simple, i.e., the Jacobian of the averaged function in it is zero, the classical averaging theory does not provide information about the periodic solution associated to a non-simple zero. Here we provide sufficient conditions in order that the averaging theory can be applied also to non-simple zeros for studying their associated periodic solutions. Additionally, we do two applications of this new result for studying the zero-Hopf bifurcation in the Lorenz system and in the Fitzhugh-Nagumo system.

  15. Pressurized Heavy Water Reactor Fuel: Integrity, Performance and Advanced Concepts. Proceedings of the Technical Meetings held in Bucharest, 24-27 September 2012, and in Mumbai, 8-11 May 2013

    International Nuclear Information System (INIS)

    Seven Member States have operating pressurized heavy water reactors (PHWRs), and some of them are also planning new reactors of this type. The current type of PHWR uses natural uranium as the fuel and has an average burnup of 7000 MWd/t (megawatt days per metric tonne). To make these reactors economically competitive with other reactor types, the discharge burnup of PHWR fuel will need to be increased without affecting the integrity of the fuel pin and bundle. A significant increase in the discharge burnup of fuel is possible with the use of advanced fuel cycles in PHWRs. The advanced fuels can be slightly enriched uranium, reprocessed uranium from light water reactors, mixed oxide or thorium based fuels. At the same time, substantial savings in natural uranium resources can also be achieved through the possible extension of the discharge burnup of advanced fuels used in PHWRs without changing reactor hardware. Following the recommendation of the Technical Working Group on Fuel Performance and Technology, two technical meetings were held: Technical Meeting on Fuel Integrity during Normal Operation and Accident Conditions in PHWRs, 24–27 September 2012, Bucharest, Romania; and Technical Meeting on Advanced Fuel Cycles in PHWRs, 8–11 April 2013, Mumbai, India. Their objective was to update information on the performance of PHWR fuels, the status and trends in the use of advanced fuels in PHWRs and the technical readiness for the deployment of such fuel cycles in these types of reactor. This publication contains the proceedings of the two technical meetings, including a record of the discussions held during the various technical sessions

  16. Short-Term Auditory Memory of Above-Average and Below-Average Grade Three Readers.

    Science.gov (United States)

    Caruk, Joan Marie

    To determine if performance on short term auditory memory tasks is influenced by reading ability or sex differences, 62 third grade reading students (16 above average boys, 16 above average girls, 16 below average boys, and 14 below average girls) were administered four memory tests--memory for consonant names, memory for words, memory for…

  17. Platinum Porous Electrodes for Fuel Cells

    DEFF Research Database (Denmark)

    Andersen, Shuang Ma

    Fuel cell energy bears the merits of renewability, cleanness and high efficiency. Proton Exchange Membrane Fuel Cell (PEMFC) is one of the most promising candidates as the power source in the near future. A fine management of different transports and electrochemical reactions in PEM fuel cells...... to a genuine picture of a working PEM fuel cell catalyst layer. These, in turn, enrich the knowledge of Three-Phase-Boundary, provide efficient tool for the electrode selection and eventually will contribute the advancement of PEMFC technology....

  18. Clarifying the Relationship between Average Excesses and Average Effects of Allele Substitutions.

    Science.gov (United States)

    Alvarez-Castro, José M; Yang, Rong-Cai

    2012-01-01

    Fisher's concepts of average effects and average excesses are at the core of the quantitative genetics theory. Their meaning and relationship have regularly been discussed and clarified. Here we develop a generalized set of one locus two-allele orthogonal contrasts for average excesses and average effects, based on the concept of the effective gene content of alleles. Our developments help understand the average excesses of alleles for the biallelic case. We dissect how average excesses relate to the average effects and to the decomposition of the genetic variance. PMID:22509178

  19. Clarifying the relationship between average excesses and average effects of allele substitutions

    Directory of Open Access Journals (Sweden)

    Jose M eÁlvarez-Castro

    2012-03-01

    Full Text Available Fisher’s concepts of average effects and average excesses are at the core of the quantitative genetics theory. Their meaning and relationship have regularly been discussed and clarified. Here we develop a generalized set of one-locus two-allele orthogonal contrasts for average excesses and average effects, based on the concept of the effective gene content of alleles. Our developments help understand the average excesses of alleles for the biallelic case. We dissect how average excesses relate to the average effects and to the decomposition of the genetic variance.

  20. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  1. Enrichment of ventilation air methane (VAM) with carbon fiber composites.

    Science.gov (United States)

    Bae, Jun-Seok; Su, Shi; Yu, Xin Xiang

    2014-05-20

    Treatment of ventilation air methane (VAM) with cost-effective technologies has been an ongoing challenge due to its high volumetric flow rate with low and variable methane concentrations. In this work, honeycomb monolithic carbon fiber composites were developed and employed to capture VAM with a large-scale test unit at various conditions such as VAM concentration, ventilation air (VA) flow rate, temperature, and purging fluids. Regardless of inlet VAM concentrations, methane was captured at almost 100%. To regenerate the composites, the initial vacuum swing followed by combined temperature and vacuum swing adsorption (TVSA) was applied. It was found that initial vacuum swing is a control step for the final methane concentration having 5 or 11 times the VAM enrichment by one-step adsorption, which is, to our knowledge, the best performance achieved in VAM enrichment technologies worldwide. Five-time enriched VAM can be utilized as a principle fuel for lean burn turbine. Also, it can be further enriched by second step adsorption to more than 25% which then can be used for commercially available gas engines. In this way, the final product can be out of the methane explosive range (5-15%). PMID:24787090

  2. Uranium enrichment: a competitive market in the future?

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre Ferreira; Honaiser, Eduardo Henrique Rangel [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)]. E-mail: 20-1@ctemsp.mar.mil.br

    2005-07-01

    Uranium enrichment is the costly step in the nuclear fuel cycle. It has born as a an activity for the military in the 40s, financed by governments, such as the United States (US) and the former Soviet Union. Later, other major nations have joined them in the nuclear weapons development. The activity of enrichment was done in each country that developed nuclear weapons, and the nuclear weapons countries, especially the US and Soviet Union, dictated the mined uranium market. In the 70s, with the growth of the commercial use of nuclear energy, uranium enrichment started to be treated as a market, which gradually have structured itself, strongly influenced by the historical background. Today, the market is an oligopoly of four major government-owned (or government-influenced) companies. In this paper, the trends in the enrichment market are identified, focusing on competitiveness. Through the conduction of a market analysis (past and future), and the study of the market structure evolution, a more competitive market is shown, but still influenced by the governmental participation. Competitiveness is dictated by government support, verticalization capacity, and, mainly by technological advantages. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    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.

  4. Small scale magnetic flux-averaged magnetohydrodynamics

    International Nuclear Information System (INIS)

    By relaxing exact magnetic flux conservation below a scale λ a system of flux-averaged magnetohydrodynamic equations are derived from Hamilton's principle with modified constraints. An energy principle can be derived from the linearized averaged system because the total system energy is conserved. This energy principle is employed to treat the resistive tearing instability and the exact growth rate is recovered when λ is identified with the resistive skin depth. A necessary and sufficient stability criteria of the tearing instability with line tying at the ends for solar coronal loops is also obtained. The method is extended to both spatial and temporal averaging in Hamilton's principle. The resulting system of equations not only allows flux reconnection but introduces irreversibility for appropriate choice of the averaging function. Except for boundary contributions which are modified by the time averaging process total energy and momentum are conserved over times much longer than the averaging time τ but not for less than τ. These modified boundary contributions correspond to the existence, also, of damped waves and shock waves in this theory. Time and space averaging is applied to electron magnetohydrodynamics and in one-dimensional geometry predicts solitons and shocks in different limits

  5. Fuel cycles using adulterated plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Brooksbank, R. E.; Bigelow, J. E.; Campbell, D. O.; Kitts, F. G.; Lindauer, R. B.

    1978-01-01

    Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with /sup 238/U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial.

  6. Experimental Investigation of Oxygen Enriched air intake on Combustion Parameters of a Single Cylinder Diesel Engine

    OpenAIRE

    Rajkumar, K; Govindarajan, P

    2010-01-01

    In the present experimental work a computerized Single cylinder Diesel engine with data acquisition system was used to study the effects of oxygen enriched air intake on combustion parameters. Increasing the oxygen content with the air leads to faster burn rates and the ability to burn more fuel at the same stoichiometery. Addedoxygen in the combustion air leads to shorter ignition delays and offers more potential for burning diesel. Oxy-fuel combustion reduces the volume of flue gases and re...

  7. Averaged Lema\\^itre-Tolman-Bondi dynamics

    CERN Document Server

    Isidro, Eddy G Chirinos; Piattella, Oliver F; Zimdahl, Winfried

    2016-01-01

    We consider cosmological backreaction effects in Buchert's averaging formalism on the basis of an explicit solution of the Lema\\^itre-Tolman-Bondi (LTB) dynamics which is linear in the LTB curvature parameter and has an inhomogeneous bang time. The volume Hubble rate is found in terms of the volume scale factor which represents a derivation of the simplest phenomenological solution of Buchert's equations in which the fractional densities corresponding to average curvature and kinematic backreaction are explicitly determined by the parameters of the underlying LTB solution at the boundary of the averaging volume. This configuration represents an exactly solvable toy model but it does not adequately describe our "real" Universe.

  8. Average-passage flow model development

    Science.gov (United States)

    Adamczyk, John J.; Celestina, Mark L.; Beach, Tim A.; Kirtley, Kevin; Barnett, Mark

    1989-01-01

    A 3-D model was developed for simulating multistage turbomachinery flows using supercomputers. This average passage flow model described the time averaged flow field within a typical passage of a bladed wheel within a multistage configuration. To date, a number of inviscid simulations were executed to assess the resolution capabilities of the model. Recently, the viscous terms associated with the average passage model were incorporated into the inviscid computer code along with an algebraic turbulence model. A simulation of a stage-and-one-half, low speed turbine was executed. The results of this simulation, including a comparison with experimental data, is discussed.

  9. Average Shape of Transport-Limited Aggregates

    Science.gov (United States)

    Davidovitch, Benny; Choi, Jaehyuk; Bazant, Martin Z.

    2005-08-01

    We study the relation between stochastic and continuous transport-limited growth models. We derive a nonlinear integro-differential equation for the average shape of stochastic aggregates, whose mean-field approximation is the corresponding continuous equation. Focusing on the advection-diffusion-limited aggregation (ADLA) model, we show that the average shape of the stochastic growth is similar, but not identical, to the corresponding continuous dynamics. Similar results should apply to DLA, thus explaining the known discrepancies between average DLA shapes and viscous fingers in a channel geometry.

  10. Averaging of Backscatter Intensities in Compounds

    Science.gov (United States)

    Donovan, John J.; Pingitore, Nicholas E.; Westphal, Andrew J.

    2002-01-01

    Low uncertainty measurements on pure element stable isotope pairs demonstrate that mass has no influence on the backscattering of electrons at typical electron microprobe energies. The traditional prediction of average backscatter intensities in compounds using elemental mass fractions is improperly grounded in mass and thus has no physical basis. We propose an alternative model to mass fraction averaging, based of the number of electrons or protons, termed “electron fraction,” which predicts backscatter yield better than mass fraction averaging. PMID:27446752

  11. Experimental Demonstration of Squeezed State Quantum Averaging

    CERN Document Server

    Lassen, Mikael; Sabuncu, Metin; Filip, Radim; Andersen, Ulrik L

    2010-01-01

    We propose and experimentally demonstrate a universal quantum averaging process implementing the harmonic mean of quadrature variances. The harmonic mean protocol can be used to efficiently stabilize a set of fragile squeezed light sources with statistically fluctuating noise levels. The averaged variances are prepared probabilistically by means of linear optical interference and measurement induced conditioning. We verify that the implemented harmonic mean outperforms the standard arithmetic mean strategy. The effect of quantum averaging is experimentally tested both for uncorrelated and partially correlated noise sources with sub-Poissonian shot noise or super-Poissonian shot noise characteristics.

  12. Self-averaging characteristics of spectral fluctuations

    OpenAIRE

    Braun, Petr; Haake, Fritz

    2014-01-01

    The spectral form factor as well as the two-point correlator of the density of (quasi-)energy levels of individual quantum dynamics are not self-averaging. Only suitable smoothing turns them into useful characteristics of spectra. We present numerical data for a fully chaotic kicked top, employing two types of smoothing: one involves primitives of the spectral correlator, the second a small imaginary part of the quasi-energy. Self-averaging universal (like the CUE average) behavior is found f...

  13. Changing mortality and average cohort life expectancy

    DEFF Research Database (Denmark)

    Schoen, Robert; Canudas-Romo, Vladimir

    2005-01-01

    of survivorship. An alternative aggregate measure of period mortality which has been seen as less sensitive to period changes, the cross-sectional average length of life (CAL) has been proposed as an alternative, but has received only limited empirical or analytical examination. Here, we introduce a new measure......, the average cohort life expectancy (ACLE), to provide a precise measure of the average length of life of cohorts alive at a given time. To compare the performance of ACLE with CAL and with period and cohort life expectancy, we first use population models with changing mortality. Then the four aggregate...

  14. Behavior analysis of U3Si-Al fuel in MP type fuel elements under irradiation

    International Nuclear Information System (INIS)

    Uranium silicide U3Si is considered as perspective nuclear fuel for Russian research reactors. In order to resolve the problem of enrichment reduction this nuclear fuel is the most real alternative for the Uranium dioxide which is currently used for these purposes. Within RERTR program two MP type fuel element models with the core consisting of U3Si nuclear fuel dispersed in an aluminium matrix were tested in MP reactor. The tests confirmed that the use of U3Si + Al fuel composition is a perspective solution to reduce fuel element enrichment in research reactors. This report represents analysis of post-irradiation tests of the fuel element models. The goal of the analysis being to establish the value and the appropriateness of swelling for the Uranium silicide. The fuel element represents a cylinder tube with four ribs on the outer surface. The claddings are produced of CAB-6 alloy. The contents of nuclear fuel in the core constitute 34% by volume, technological pores constitute 4.5% and the rest is aluminium matrix. The nuclear fuel was produced in ARSRIIM, the fuel elements was produced by ARSRIIM specialists with equipment of NZKH. (author)

  15. Data mining in the study of nuclear fuel cells; Mineria de datos en el estudio de celdas de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Medina P, J. A. [Universidad Autonoma de Campeche, Av. Agustin Melgar s/n, Col. Buenavista, 24039 San Francisco de Campeche, Campeche (Mexico); Ortiz S, J. J.; Castillo, A.; Montes T, J. L.; Perusquia, R., E-mail: j.angel.mp@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    In this paper is presented a study of data mining application in the analysis of fuel cells and their performance within a nuclear boiling water reactor. A decision tree was used to fulfill questions of the type If (condition) and Then (conclusion) to classify if the fuel cells will have good performance. The performance is measured by compliance or not of the cold shutdown margin, the rate of linear heat generation and the average heat generation in a plane of the reactor. It is assumed that the fuel cells are simulated in the reactor under a fuel reload and rod control patterns pre designed. 18125 fuel cells were simulated according to a steady-state calculation. The decision tree works on a target variable which is one of the three mentioned before. To analyze this objective, the decision tree works with a set of attribute variables. In this case, the attributes are characteristics of the cell as number of gadolinium rods, rods number with certain uranium enrichment mixed with a concentration of gadolinium, etc. The found model was able to predict the execution or not of the shutdown margin with a precision of around 95%. However, the other two variables showed lower percentages due to few learning cases of the model in which these variables were or were not achieved. Even with this inconvenience, the model is quite reliable and can be used in way coupled in optimization systems of fuel cells. (Author)

  16. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

    Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO2, lower sensitivity to 234U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

  17. Job enrichment in job design.

    Science.gov (United States)

    Bobeng, B J

    1977-03-01

    For optimal operation in labor-intensive industries, such as foodservice, not only scientific management principles but also behavioral aspects (the people) must be considered in designing job content. Three psychologic states--work that is meaningful, responsibility for outcomes, and knowledge of outcomes--are critical in motivating people. These, in turn encompass the core dimensions of skill variety, task identity, task significance, autonomy, and feedback. Job enrichment and job enlargement--related but not identical means of expanding job content--when combined, offer the likelihood of redesigned jobs in the core dimensions. Effective implementation of a job enrichment program hinges on diagnosing problems in the work system, actual changes in the work, and systematic evaluation of the changes. The importance of the contribution of the behavioral sciences to management cannot be neglected.

  18. Average Vegetation Growth 1992 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1992 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  19. Average Vegetation Growth 1994 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1994 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  20. Average Vegetation Growth 1991 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1991 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  1. Average Vegetation Growth 1993 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1993 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  2. Average Vegetation Growth 1998 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1998 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  3. Average Vegetation Growth 1999 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1999 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  4. Average Vegetation Growth 1990 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1990 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  5. Average Vegetation Growth 2003 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 2003 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  6. Sea Surface Temperature Average_SST_Master

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Sea surface temperature collected via satellite imagery from http://www.esrl.noaa.gov/psd/data/gridded/data.noaa.ersst.html and averaged for each region using...

  7. MN Temperature Average (1961-1990) - Polygon

    Data.gov (United States)

    Minnesota Department of Natural Resources — This data set depicts 30-year averages (1961-1990) of monthly and annual temperatures for Minnesota. Isolines and regions were created using kriging and...

  8. Average Vegetation Growth 2002 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 2002 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  9. Average Vegetation Growth 1997 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1997 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  10. Spacetime Average Density (SAD) Cosmological Measures

    CERN Document Server

    Page, Don N

    2014-01-01

    The measure problem of cosmology is how to obtain normalized probabilities of observations from the quantum state of the universe. This is particularly a problem when eternal inflation leads to a universe of unbounded size so that there are apparently infinitely many realizations or occurrences of observations of each of many different kinds or types, making the ratios ambiguous. There is also the danger of domination by Boltzmann Brains. Here two new Spacetime Average Density (SAD) measures are proposed, Maximal Average Density (MAD) and Biased Average Density (BAD), for getting a finite number of observation occurrences by using properties of the Spacetime Average Density (SAD) of observation occurrences to restrict to finite regions of spacetimes that have a preferred beginning or bounce hypersurface. These measures avoid Boltzmann brain domination and appear to give results consistent with other observations that are problematic for other widely used measures, such as the observation of a positive cosmolo...

  11. Averaging procedure in variable-G cosmologies

    CERN Document Server

    Cardone, Vincenzo F

    2008-01-01

    Previous work in the literature had built a formalism for spatially averaged equations for the scale factor, giving rise to an averaged Raychaudhuri equation and averaged Hamiltonian constraint, which involve a backreaction source term. The present paper extends these equations to include models with variable Newton parameter and variable cosmological term, motivated by the non-perturbative renormalization program for quantum gravity based upon the Einstein--Hilbert action. The coupling between backreaction and spatially averaged three-dimensional scalar curvature is found to survive, and all equations involving contributions of a variable Newton parameter are worked out in detail. Interestingly, under suitable assumptions, an approximate solution can be found where the universe tends to a FLRW model, while keeping track of the original inhomogeneities through two effective fluids.

  12. A practical guide to averaging functions

    CERN Document Server

    Beliakov, Gleb; Calvo Sánchez, Tomasa

    2016-01-01

    This book offers an easy-to-use and practice-oriented reference guide to mathematical averages. It presents different ways of aggregating input values given on a numerical scale, and of choosing and/or constructing aggregating functions for specific applications. Building on a previous monograph by Beliakov et al. published by Springer in 2007, it outlines new aggregation methods developed in the interim, with a special focus on the topic of averaging aggregation functions. It examines recent advances in the field, such as aggregation on lattices, penalty-based aggregation and weakly monotone averaging, and extends many of the already existing methods, such as: ordered weighted averaging (OWA), fuzzy integrals and mixture functions. A substantial mathematical background is not called for, as all the relevant mathematical notions are explained here and reported on together with a wealth of graphical illustrations of distinct families of aggregation functions. The authors mainly focus on practical applications ...

  13. MN Temperature Average (1961-1990) - Line

    Data.gov (United States)

    Minnesota Department of Natural Resources — This data set depicts 30-year averages (1961-1990) of monthly and annual temperatures for Minnesota. Isolines and regions were created using kriging and...

  14. Average Vegetation Growth 2001 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 2001 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  15. Monthly snow/ice averages (ISCCP)

    Data.gov (United States)

    National Aeronautics and Space Administration — September Arctic sea ice is now declining at a rate of 11.5 percent per decade, relative to the 1979 to 2000 average. Data from NASA show that the land ice sheets...

  16. Boron Enrichment in Martian Clay

    OpenAIRE

    James D Stephenson; Lydia J Hallis; Kazuhide Nagashima; Freeland, Stephen J.

    2013-01-01

    We have detected a concentration of boron in martian clay far in excess of that in any previously reported extra-terrestrial object. This enrichment indicates that the chemistry necessary for the formation of ribose, a key component of RNA, could have existed on Mars since the formation of early clay deposits, contemporary to the emergence of life on Earth. Given the greater similarity of Earth and Mars early in their geological history, and the extensive disruption of Earth's earliest minera...

  17. Enrichment of lanthanides in aragonite

    Institute of Scientific and Technical Information of China (English)

    瞿成利; 路波; 刘刚

    2009-01-01

    Using the constant addition technique,the coprecipitation of lanthanum,gadolinium,and lutetium with aragonite in seawater was experimentally investigated at 25 ℃.Their concentrations in aragonite overgrowths were determined by inductive coupled plasma mass spectrometer.All these lanthanides were strongly enriched in aragonite overgrowths.The amount of lanthanum,gadolinium,and lutetium incorporated into aragonite accounted for 57%-99%,50%-89%,and 40%-91% of their initial total amount,respectively.With the in...

  18. Modeling and Instability of Average Current Control

    OpenAIRE

    Fang, Chung-Chieh

    2012-01-01

    Dynamics and stability of average current control of DC-DC converters are analyzed by sampled-data modeling. Orbital stability is studied and it is found unrelated to the ripple size of the orbit. Compared with the averaged modeling, the sampled-data modeling is more accurate and systematic. An unstable range of compensator pole is found by simulations, and is predicted by sampled-data modeling and harmonic balance modeling.

  19. Nonequilibrium statistical averages and thermo field dynamics

    International Nuclear Information System (INIS)

    An extension of thermo field dynamics is proposed, which permits the computation of nonequilibrium statistical averages. The Brownian motion of a quantum oscillator is treated as an example. In conclusion it is pointed out that the procedure proposed to computation of time-dependent statistical average gives the correct two-point Green function for the damped oscillator. A simple extension can be used to compute two-point Green functions of free particles

  20. Average Bandwidth Allocation Model of WFQ

    Directory of Open Access Journals (Sweden)

    Tomáš Balogh

    2012-01-01

    Full Text Available We present a new iterative method for the calculation of average bandwidth assignment to traffic flows using a WFQ scheduler in IP based NGN networks. The bandwidth assignment calculation is based on the link speed, assigned weights, arrival rate, and average packet length or input rate of the traffic flows. We prove the model outcome with examples and simulation results using NS2 simulator.