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Sample records for reduced enrichment fuels

  1. Comments on applications of reduced enrichment fuels

    International Nuclear Information System (INIS)

    Winkler, M.H.

    1983-01-01

    Full text: I will briefly describe the experience gained using different fuels in the SAPHIR reactor in Switzerland. The SAPHIR has been operating since 1957 and was the first swimming pool reactor built outside of the United States, which was originally known as the Geneva Conference Reactor. The first core was loaded with 20 percent enriched high density UO 2 fuel with a density of about 2.5 grams per cc, fabricated in 1955 by Oak Ridge National Laboratory. After a few years of operation at a power level of one MW, more than one batch of the elements released small amounts of fission products mainly Xe and Kr. When these releases were discovered, high enriched fuel was becoming available so that the fuel fabricators began to produce the lower density high enriched fuels. During this transition from fabrication of low to high enriched fuels no one could foresee that the stone age of nuclear fuel fabrication would come back again. Therefore, we did not investigate the reasons for the fission product release from the high density low enriched UO 2 fuel. The second fuel type used in the SAPHIR was the 90 percent enriched low density U 3 O 8 fuel fabricated by NUKEM. This high enriched fuel has performed satisfactorily over the years. Since 1968, the core has been using improved 23 plate fuel elements with a loading of 280 grams of uranium. The reactor power has been recently increased to five MW. An additional increase in the power level to 10 MW is planned at the end of next year so that heavier loaded elements will be needed. In order to follow the recommendations of the INFCE working group 8C and in cooperation with the reduced enrichment program, we intend to initially reduce the fuel enrichment to 45 percent. Last year we ordered five fuel elements with a loading of 320 grams 235 U/element and 45 percent enrichment for full power tests. Unfortunately, the delivery of the necessary enriched fuel uranium has been delayed and it is not available at this time. If

  2. Reducing enrichment of fuel for research reactors

    International Nuclear Information System (INIS)

    Kanda, Keiji; Matsuura, Shojiro.

    1980-01-01

    In research reactors, highly enriched uranium (HEU) is used as fuel for their purposes of operation. However, the United States strongly required in 1977 that these HEU should be replaced by low enrichment uranium (LEU) of 20% or less, or even in unavoidable cases, it should be replaced by medium enrichment uranium (MEU). INFCE (International Nuclear Fuel Cycle Evaluation) which started its activity just at that time decided to discuss this problem in the research reactor group of No. 8 sectional committee. Japan has been able to forward the work, taking a leading part in the international opinion because she has taken the countermeasures quickly. INFCE investigated the problem along the lines of policy that the possibility of reducing the degree of enrichment should be limited to the degree in which the core structures and equipments of research reactors will be modified as little as possible, and the change of fuel element geometry will be done within the permissible thermohydrodynamic capacity, and concluded that it might be possible in near future to reduce the degree of enrichment to about 45% MEU, while the reduction to 20% LEU might require considerable research, development and verification. On the other hand, the joint researches by Kyoto University and ANL (Argonne National Laboratory) and by Japan Atomic Energy Research Institute and ANL are being continued. IAEA has edited the guidebook (IAEA-TECDOC-233) for reducing the degree of enrichment for developing countries. (Wakatsuki, Y.)

  3. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

    Snelgrove, James L.

    1983-01-01

    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

  4. Reduced enriched fuel status at CERCA

    International Nuclear Information System (INIS)

    Tissier, A.; Fanjas, Y.

    1991-01-01

    CERCA's main objective is to satisfy its customers, improving quality of its products, and maintaining the costs as low as possible. Its Research and Development program reveals this goal. Different R and D topics under development at short (recycling of scraps), at medium (X-ray imaging machine) and at long term (improvement of fuel materials) are presented as evidence of this will. (orig.)

  5. Effect of reduced enrichment on the fuel cycle for research reactors

    International Nuclear Information System (INIS)

    Travelli, A.

    1982-01-01

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel

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

    International Nuclear Information System (INIS)

    1983-08-01

    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

  7. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wood, J C; Foo, M T; Berthiaume, L C; Herbert, L N; Schaefer, J D; Hawley, D [Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, ON KOJ 1JO (Canada)

    1985-07-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U{sub 3}Si in aluminum, to complement the dispersions of U{sub 3}Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U{sub 3}Si have been manufactured. (author)

  8. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.; Hawley, D.

    1985-01-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U 3 Si in aluminum, to complement the dispersions of U 3 Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U 3 Si have been manufactured. (author)

  9. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    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

  10. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Shibata, Toshikazu.

    1982-01-01

    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 U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  12. Central fuel banking to reduce the number of proliferation sensitive enrichment activities

    International Nuclear Information System (INIS)

    Cserhati, A.

    2008-01-01

    Central fuel banking is a complex international political, economic and technical concept that aims to reduce uncontrolled spreading of uranium enrichment technology in the world in order to prevent proliferation of nuclear weapons. This paper first gives an outline of the notions: 'non-proliferation', the 'front-end' of the fuel cycle, the scope of fuel baking, nuclear fuel and the 60 years of enrichment technology. Enrichment technology is highly concentrated in the nuclear weapon states and other developed countries, but this is not exclusive any more. The technology is spreading. The global demand for enrichment services - parallel to massive nuclear investments in the civil sector and the ageing of older facilities - is constantly growing. Proliferation sensitivity calls for an effective and comprehensive non-proliferation regime. The solution may be multilateralizing the nuclear fuel cycle. After a historical overview, the proposals on multilateral nuclear approaches are presented. The assessment of the proposals is complex in the dimensions of: the non-proliferation aim, the assurance of supply aspect and other variables such as legal issues and non-nuclear inducements. A general evaluation and the recommendations of the Expert Panel of the IAEA are introduced outlining a plan on a middle- and long-term basis. The conclusion of the paper stresses the importance and challenge in finding the 'new balance' between obligations and interests of the members of the global community stating that the answers will have a significant impact on the nuclear indus- try, world wide economics and security policy. (orig.)

  13. Thermal breeder fuel enrichment zoning

    International Nuclear Information System (INIS)

    Capossela, H.J.; Dwyer, J.R.; Luce, R.G.; McCoy, D.F.; Merriman, F.C.

    1992-01-01

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

  14. Reduced enrichment fuel and its reactivity effects in the University Training Reactor Moata

    International Nuclear Information System (INIS)

    Wilson, D.J.

    1983-08-01

    Concern for nuclear proliferation is likely to preclude future supply of highly enriched uranium fuel for research reactors such as the University Training Reactor Moata. This study calculates the fuel densities necessary to maintain the reactivity per plate of the present high enrichment (90 per cent 235 U) fuel for a range of lower enrichments assuming that no geometry changes are allowed. The maximum uranium density for commercially available aluminium-type research reactor fuels is generally considered to be about 1.7 g cm -3 . With this density limitation, the minimum enrichment to maintain present reactivity per plate is about 35 per cent 235 U. For low enrichment (max. 20 per cent 235 U) fuel, the required U density is about 2.9 g cm -3 , which is beyond the expected range for UAl/sub x/-Al but within that projected for the longer term development and full qualification for U 3 O 8 -Al. Medium enrichment (nominally 45 per cent 235 U) Al/sub x/-Al would be entirely satisfactory as an immediate replacement fuel, requiring no modifications to the reactor and operating procedures, and minimal reappraisal of safety issues. Included in this study are calculations of the fuel coefficients at various enrichments, the effect of replacing standard fuel plates or complete elements with 45 per cent enriched fuel, and the reactivity to be gained by replacing 12-plate with 13-plate elements

  15. 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.

  16. The reduced enrichment program for JRR-4

    International Nuclear Information System (INIS)

    Takayanagi, M.

    1992-01-01

    Japan Research Reactor No. 4(JRR-4) with the rated power of 3.5 MW, swimming pool type research reactor, 93 % enriched uranium ETR-type fuel used, light water moderated and cooled. The first criticality reached on 28th January, 1965. The reactor has operated for about 26 years. However, it was planed to the reduced enrichment of the fuels to low enrichment according to the International Reduced Enrichment for Research and Test Reactors (RERTR) program. This paper describes the program for conversion of the enrichment of fuel from 93 % to less than 20 %. (author)

  17. Study of the reduced enrichment fuel conversion at the University of Missouri-Rolla reactor

    International Nuclear Information System (INIS)

    Straka, M.; Bolon, A.; Covington, L.

    1987-01-01

    The method used to analyze the low-enriched uranium core which has been proposed for the University of Missouri-Rolla Reactor is described. Results of calculations for the high-enriched uranium core have been compared with the measured data whenever possible in order to verify this method. For most of the cases that were analyzed the proposed method is adequate and the results obtained for the low-enriched uranium core can be used in revising the licensing documents. (Author)

  18. 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.

  19. The development and testing of reduced enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.

    1983-01-01

    Fuel rods of uranium silicide dispersed in aluminum and clad in aluminum have been developed and tested in the laboratory and in-reactor. The properties of the dispersion fuel materials proved satisfactory with regard to thermal conductivity, aqueous corrosion resistance, strength and ductility, and thermal stability below 473 K. A vacancy condensation model is proposed to account for the thermally-induced swelling that occurs above 473 K by virtue of the chemical reactions that occur between the dispersed silicide fuel particles and the aluminum matrix. The in-reactor fuel core swelling was less than % after irradiation at high powers 76-131 kW/m) to a high terminal burnup (79.2 at% of U-235 atoms). (author)

  20. Advances in the manufacturing and irradiation of reduced enrichment fuels for canadian research reactors

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

    1984-01-01

    The procedures for manufacturing fuel rods of uranium silicide dispersed in aluminum and clad in aluminum have been optimized to maximize production rates while minimizing scrap losses. Melting and casting, chip machining and core extrusion have all been re-evaluated to improve their efficiency and significant gains have been made, whilst maintaining high quality standards. The results of our irradiation program on mini-elements up to a burnup of 80 atomic percent continue to be encouraging. The upper bound curve of fuel core swelling versus burnup in the range 0-80 atomic percent represents 1% swelling per 10 atomic percent burnup. Fuel core swelling has now been measured directly on six mini-elements from which the clad surface oxide had been removed showing that previous calculated values of core swelling were marginally conservative. (author)

  1. PWR fuel of high enrichment with erbia and enriched gadolinia

    International Nuclear Information System (INIS)

    Bejmer, Klaes-Håkan; Malm, Christian

    2011-01-01

    Today standard PWR fuel is licensed for operation up to 65-70 MWd/kgU, which in most cases corresponds to an enrichment of more than 5 w/o "2"3"5U. Due to criticality safety reason of storage and transportation, only fuel up to 5 w/o "2"3"5U enrichment is so far used. New fuel storage installations and transportation casks are necessary investments before the reactivity level of the fresh fuel can be significantly increased. These investments and corresponding licensing work takes time, and in the meantime a solution that requires burnable poisons in all pellets of the fresh high-enriched fuel might be used. By using very small amounts of burnable absorber in every pellet the initial reactivity can be reduced to today's levels. This study presents core calculations with fuel assemblies enriched to almost 6 w/o "2"3"5U mixed with a small amount of erbia. Some of the assemblies also contain gadolinia. The results are compared to a reference case containing assemblies with 4.95 w/o "2"3"5U without erbia, utilizing only gadolinia as burnable poison. The comparison shows that the number of fresh fuel assemblies can be reduced by 21% (which increases the batch burnup by 24%) by utilizing the erbia fuel concept. However, increased cost of uranium due to higher enrichment is not fully compensated for by the cost gain due to the reduction of the number assemblies. Hence, the fuel cycle cost becomes slightly higher for the high enrichment erbia case than for the reference case. (author)

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

    International Nuclear Information System (INIS)

    Hickey, B.

    1983-01-01

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

  3. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    Aguilar H, F.; Mazon R, R.

    2001-05-01

    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)

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

    International Nuclear Information System (INIS)

    Lee, Duu-Jong; Lee, Chin-Yu; Chang, Jo-Shu

    2012-01-01

    Highlights: ► We started up microbial fuel cell (MFC) using enriched sulfate-reducing mixed culture. ► Sulfate-reducing bacteria and anode-respiring bacteria were enriched in anodic biofilms. ► The MFC effectively remove sulfate to elementary sulfur in the presence of lactate. ► 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.

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-08-01

    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.

  7. 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)

    1983-08-01

    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

  8. Development of long-life low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.J.; West, G.B.

    1978-01-01

    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

  9. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance and fuel particle chemical performande. (orig.) [de

  10. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance

  11. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-07-01

    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

  12. 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.

  13. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  14. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    Stahl, D.

    1993-01-01

    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

  15. Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

    International Nuclear Information System (INIS)

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

    1984-01-01

    According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

  16. Slightly enriched uranium fuel for a PHWR

    International Nuclear Information System (INIS)

    Notari, C.; Marajofsky, A.

    1997-01-01

    An improved fuel element design for a PHWR using slightly enriched uranium fuel is presented. It maintains the general geometric disposition of the currently used in the argentine NPP's reactors, replacing the outer ring of rods by rods containing annular pellets. Power density reduction is achieved with modest burnup losses and the void volume in the pellets can be used to balance these two opposite effects. The results show that with this new design, the fuel can be operated at higher powers without violating thermohydraulic limits and this means an improvement in fuel management flexibility, particularly in the transition from natural uranium to slightly enriched uranium cycle. (author)

  17. From high enriched to low enriched uranium fuel in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L. [Nuclear Materials Science Institute, SCK.CEN, Boeretang 200, B-2400 Mol (Belgium)

    2010-07-01

    Since the 1970's, global efforts have been going on to replace the high-enriched (>90% {sup 235}U), low-density UAlx research reactor fuel with high-density, low enriched (<20% {sup 235}U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U{sub 3}Si{sub 2} dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U{sub 3}Si{sub 2} (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

  18. From high enriched to low enriched uranium fuel in research reactors

    International Nuclear Information System (INIS)

    Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L.

    2010-01-01

    Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235 U), low-density UAlx research reactor fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U 3 Si 2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U 3 Si 2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

  19. Experience with a fuel rod enrichment scanner

    International Nuclear Information System (INIS)

    Kubik, R.N.; Pettus, W.G.

    1975-01-01

    This enrichment scanner views all fuel rods produced at B and W's Commercial Nuclear Fuel Plant. The scanner design is derived from the PAPAS System reported by R. A. Forster, H. D. Menlove, and their associates at Los Alamos. The spatial resolution of the system and smoothing of the data are discussed in detail. The cost-effectiveness of multi-detector versus single detector scanners of this general design is also discussed

  20. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    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 3 was by then in routine use, illustrated how far work has progressed

  1. 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.

  2. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  3. Reduction of fuel enrichment for research reactors built-up in accordance with Russian (Soviet) projects

    International Nuclear Information System (INIS)

    Aleksandrov, A.B.; Enin, A.A.; Tkachyov, A.A.

    2001-01-01

    In accordance with the Russian program of reduced enrichment for research and test reactors (RERTR) built-up in accordance with Russian (Soviet) projects, AO 'NCCP' performs works on FA fabrication with reduced enrichment fuel. The main trends and results of performed works on research reactors FEs and FAs based on UO 2 and U-9%Mo fuel with U 235 19.7% enrichment are described. (author)

  4. Status of reduced enrichment program for research reactors in Japan

    International Nuclear Information System (INIS)

    Kaieda, Keisuke; Baba, Osamu; Nagaoka, Yoshiharu; Kanda, Keiji; Nakagome, Yoshihiro

    1999-01-01

    The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI) have been completed. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M commenced using LEU silicide fuel elements instead of LEU aluminide fuel elements in September, 1999. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and April 1994 the U.S. Government gave an approval to utilize HEU fuel in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until March 2004, then the full core conversion with LEU silicide will be done. The first shipment of spent fuels since 1974 was done in August, 1999. (author)

  5. Analysis of a PHWR slightly enriched fuel

    International Nuclear Information System (INIS)

    Notari, C.; Marajofsky, A.

    1994-01-01

    It is widely known that the use of slightly enriched uranium in PHWR reactors presents economic advantages derived from the fact that less uranium is required for producing the same amount of energy. Several studies related with the use of this alternative in Atucha I NPP have been performed. The fuel assembly geometry considered up to now has been almost identical to the natural uranium one. In this work a modification consisting in the use of annular pellets in the outer ring of the cluster is analyzed. This design produces several performance benefits. The redistribution of the power in the fuel improves the maximum to average bundle power ratio. The improvement achieved depends on the void volume in the pellets which at the same time represents a certain burnup decrease. These parameters (power ratios and burnup loss) are quantified for the Atucha I and Embalse NPPs. This design improves the fuel behaviour with respect to the burnup extension derived from the slight enrichment. It is also interesting in case an overall power increase is considered. (author). 16 refs, 8 figs, 1 tab

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

    International Nuclear Information System (INIS)

    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 ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 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 99 Mo, 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 U 3 Si-Al with 19.75 % enrichment and U 3 Si 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. (Author)

  7. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  8. Reduced enrichment program for the FRM-II, status 2004

    International Nuclear Information System (INIS)

    Roehrmoser, A.; Petry, W.; Boening, K; Wieschalla, N.

    2005-01-01

    The new research reactor FRM-II of the Technische Universitaet Muenchen (TUM) has been designed to provide a maximal thermal neutron flux at mere 20 MW power. The single element design uses silicide fuel of densities 3.0 and 1.5 g/cm 3 of highly enriched uranium (HEU, 93 % U-235). With the nuclear license, that was granted in May 2003, a condition was imposed to reduce the enrichment of FRM-II to medium enriched uranium (MEU) with not more than 50 % U-235 until the end of the year 2010. The TUM has established an international working group to meet this target. This paper presents the backgrounds and the results and plannings for the first of three 2 1/2 year periods to reach the conversion in time. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    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-U{sub 3}O{sub 8} 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. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.

    1987-01-01

    The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

  11. The use of medium enriched uranium fuel for research reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The evaluation described in the present paper concerns the use of medium enriched uranium fuel for our research reactors. The underlying assumptions set up for the evaluation are as follows: (1) At first, the use of alternative fuel should not affect, even to a small extent, research and development programs in nuclear energy utilization, which were described in the previous paper. Hence the use of lower enrichment fuel should not cause any reduction in reactor performances. (2) The fuel cycle cost for operating research reactors with alternative fuel, excepting R and D cost for such fuel, should not increase beyond an acceptable limit. (3) The use of alternative fuel should be satisfactory with respect to non-proliferation purposes, to the almost same degree as the use of 20% enriched uranium fuel

  12. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

  13. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately

  14. Qualification status of LEU [low enriched uranium] fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

    Sufficient data has been obtained from tests of high-density, low-enriched fuels for research and test reactors to declare them qualified for use. These fuels include UZrH x (TRIGA fuel) and UO 2 (SPERT fuel) for rod-type reactors and UAl x , U 3 O 8 , U 3 Si 2 , and U 3 Si dispersed in aluminium for plate-type reactors. Except for U 3 Si, the allowable fission density for LEU applications is limited only by the available 235 U. Several reactors are now using these fuels, and additional conversions are in progress. The basic performance characteristics and limits, if any, of the qualified low-enriched (and medium-enriched) fuels are discussed. Continuing and planned work to qualify additional fuels is also discussed. (Author)

  15. Conversion of research reactors to low-enrichment uranium fuels

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1983-01-01

    There are at present approximately 350 research reactors in 52 countries ranging in power from less than 1 watt to 100 Megawatt and over. In the 1970's, many people became concerned about the possibility that some fuels and fuel cycles could provide an easy route to the acquisition of nuclear weapons. Since enrichment to less than 20% is internationally recognized as a fully adequate barrier to weapons usability, certain Member States have moved to minimize the international trade in highly enriched uranium and have established programmes to develop the technical means to help convert research reactors to the use of low-enrichment fuels with minimum penalties. This could involve modifications in the design of the reactor and development of new fuels. As a result of these programmes, it is expected that most research reactors can be converted to the use of low-enriched fuel

  16. Minimization of waste from uranium purification, enrichment and fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications.

  17. Minimization of waste from uranium purification, enrichment and fuel fabrication

    International Nuclear Information System (INIS)

    1999-10-01

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications

  18. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  19. Low-enriched research reactor fuel: Post-Irradiation Examinations at SCK-CEN

    International Nuclear Information System (INIS)

    Van den Berghe, S.; Leenaers, A.

    2007-01-01

    Generally, research and test reactors are fuelled with fuel plates instead of pins. In most cases in the past, these plates consisted of high enriched (higher than 95 percent 235 U) UAl 3 powder mixed with a pure Al matrix (called the meat) in between two aluminium alloy plates (the cladding). These plates are then assembled in fuel elements of different designs to fit the needs of the various reactors. Since the 1970's, efforts have been going on to replace the high-enriched, low-density UAl 3 fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched materials because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative and the Reduced Enrichment for Research and Test Reactors program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has been obtained with U 3 Si 2 fuel, which is currently used in many research reactors in the world. However, efforts to search for a better replacement have continued and are currently directed towards the U-Mo alloy fuel (7-10 weight percent Mo)

  20. Optimal pin enrichment distributions in nuclear reactor fuel bundles

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1976-01-01

    A methodology has been developed to determine the fuel pin enrichment distribution that yields the best approximation to a prescribed power distribution in nuclear reactor fuel bundles. The problem is formulated as an optimization problem in which the optimal pin enrichments minimize the sum of squared deviations between the actual and prescribed fuel pin powers. A constant average enrichment constraint is imposed to ensure that a suitable value of reactivity is present in the bundle. When constraints are added that limit the fuel pins to a few enrichment types, one must determine not only the optimal values of the enrichment types but also the optimal distribution of the enrichment types amongst the pins. A matrix of boolean variables is used to describe the assignment of enrichment types to the pins. This nonlinear mixed integer programming problem may be rigorously solved with either exhaustive enumeration or branch and bound methods using a modification of the algorithm from the continuous problem as a suboptimization. Unfortunately these methods are extremely cumbersome and computationally overwhelming. Solutions which require only a moderate computational effort are obtained by assuming that the fuel pin enrichments in this problem are ordered as in the solution to the continuous problem. Under this assumption search schemes using either exhaustive enumeration or branch and bound become computationally attractive. An adaptation of the Hooke--Jeeves pattern search technique is shown to be especially efficient

  1. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    Energy Technology Data Exchange (ETDEWEB)

    Parker, Frank L. [Vanderbilt University (United States)

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded

  2. Progress of the United States foreign research reactor spent nuclear fuel acceptance program. Reduced enrichment for research and test reactors conference 2002

    International Nuclear Information System (INIS)

    Clapper, Maureen

    2002-01-01

    Foreign Research Reactor Spent nuclear fuel Acceptance Program is actively working with research reactors to accept eligible material before the Acceptance Policy proper expires in 2006. Reactors/governments wishing to participate should contact US immediately if they have not done so already. Program operations are changing to adapt to new challenges. We continue to promote the importance of this Program to senior management in the Department of Energy

  3. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Primm, Trent; Guida, Tracey

    2010-01-01

    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.

  6. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  7. Status of reduced enrichment programs for research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

    1997-01-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE

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

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gohar, Yousry

    2011-01-01

    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)

  9. 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)

  10. 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.

  11. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, 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 HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, 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 HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  12. Refueling the RPI reactor facility with low-enrichment fuel

    International Nuclear Information System (INIS)

    Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

    1985-01-01

    The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

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

    International Nuclear Information System (INIS)

    Estryk, G.; Higa, M.

    1993-01-01

    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)

  14. Draining Water from Aircraft Fuel Using Nitrogen Enriched Air

    Directory of Open Access Journals (Sweden)

    Michael Frank

    2018-04-01

    Full Text Available This paper concerns a computational study of the process of removing water from an aircraft’s fuel tank by pumping nitrogen enriched air (NEA from the bottom of the tank. This is an important procedure for the smooth, efficient, and safe operation of the aircraft’s engine. Due to the low partial pressure of water in the pumped NEA, it absorbs water from the fuel. The water-laden bubbles enter the ullage, the empty space above the fuel, and escape into the environment. The effects of the number of NEA inlets and the NEA mass flow rate on the timescale of the NEA pumping were investigated using Computational Fluid Dynamics. The results reveal that the absorption of water by the NEA bubbles is low and is not affected by the number of the inlets used. Yet, the water content in the fuel decreases fast during the procedure, which is the desired outcome. We show that this is due to the relatively dry NEA entering the ullage and displacing the moist air, thus reducing the partial pressure of water at the fuel/ullage interface. This shift from equilibrium conditions forces water to evaporate from the fuel’s entire surface. Furthermore, the amount of water migrating from the fuel directly into the ullage is significantly greater than that absorbed by the rising bubbles. In turn, the rate of decrease of the water content in the ullage is determined by the total NEA mass flow rate and this is the dominant contributor to the draining time, with the number of NEA nozzles playing a minor role. We confirmed this by pumping NEA directly into the ullage, where we observe a significant decrease of water even when the NEA is not pumped through the fuel. We also show that doubling the mass flow rate halves the draining time. When considering the capability of most modern aircraft to pump NEA through the fuel as part of their inerting system, the proposed method for removing water is particularly attractive, requiring very little (if at all design modification.

  15. Impact of UO{sub 2} Enrichment of Fuel Zoning Rods in Long Cycle Operation of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Cheol; Lee, Deokjung [KHNP CRI, Daejeon (Korea, Republic of); Jeong, Eun; Choe, Jiwon [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    Extending the cycle length can not only increase the energy production, but also bring down outage costs by reducing the number of refueling outages during the lifetime of a nuclear power plant. It is reasonable that more fresh fuels are loaded for long cycle operation. However, minimizing the number of fresh fuels is essential in aspect of fuel economics. This can cause high power peaking near the water holes, due to increased thermalization of neutrons in those regions. To prevent this, special fuel zoning rods are used and surround the water holes. These rods use lower-enriched uranium (they have an enrichment rate lower than the other fuel rods). If we adjust the enrichment rate of fuel zoning rods, we can reduce power peaking and moreover increase cycle length. In this paper, we designed a core suitable for long cycle operation and we conducted sensitivity tests of fuel cycle length on UO2 enrichment rate in fuel zoning region in order to extend the cycle length while using the same number of fresh fuels. The correlations between the fuel zoning enrichment and cycle length, peaking factor, CBC and shutdown margin were analyzed. The more the enrichment rate in fuel zoning region increases, the more the fuel cycle length increases. At the same time, CBC, Fq and shutdown margin do not change significantly. Increasing the fuel zoning enrichment rate presents the right property of increasing the fuel cycle length without causing a large change to CBC, Fq and shutdown margin. In conclusion, by increasing the uranium enrichment rate in fuel zoning region, fuel cycle length can be increased and the safety margins can be maintained for long cycle operation of cores.

  16. Low enrichment fuel development at INEL

    International Nuclear Information System (INIS)

    Newton, D.G.

    1993-01-01

    EG and G Idaho, Inc. is under contract to the Department of Energy to operate the Idaho National Engineering Laboratory (INEL). The INEL is located in southeastern Idaho. This facility has been operating since 1949 and was originally called the National Reactor Testing Station. Several contractors manage projects on this facility. Most projects at INEL are concerned with either reactor safety or irradiation testing. At Test Area North, for example, experiments are being conducted on the effects of loss of coolant. At the Test Reactor Area the ATR (Advanced Test Reactor) and ETR (Engineering Test Reactor) are used for irradiation testing and, of course, those of you working at Argonne will recognize the Experimental Breeder Reactors I and II. SPERT is an acronym for Special Power Excursion Reactor Test. A part of this former reactor facility has been converted into a fuel fabrication laboratory facility. At SPERT IV a miniature fabrication facility has been set up to duplicate the aluminide plate fuel processing line at Atomics International. In other words, a model of the supplier's processing has been created, so that what process changes are developed here can then be scaled up to production. The process is described showing: making UAI x powder, making compact for fuel core, making experimental fuel plate and compact assembly, inspection and testing the fuel plate. Main concern was related to possible swelling

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

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J B [eds.

    1995-07-01

    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.

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

    International Nuclear Information System (INIS)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J.B.

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    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

  20. PULSTAR fuel, low enrichment, long lifetime, economical, proven

    International Nuclear Information System (INIS)

    Carter, Robert E.; Leonard, Bobby E.

    1993-01-01

    In 1962, the Western New York Research Center, Inc., located at the State University of New York at Buffalo, decided they had a need for a reactor with pulsing and high power steady state capabilities. Both General Atomic and the American Machine and Foundry Corporation (AMF) were contacted to ascertain if it were feasible to construct a dual purpose reactor of this type. The General Atomic proposal indicated the feasibility but would not warrant a steady state power of 2 MW with ultimate capability of 5 MW. AMF did provide a conceptual design for such a dual reactor, call the PULSTAR, and sufficient design information to confirm that the operating specifications could be met. The PULSTAR fuel consisted of 6 enrichment UO 2 sintered pellets in zircaloy tubes (pins) mounted in a x 5 array inside a fuel assembly. The fuel design was patterned after fuel that was under development for light water power reactors and that had been extensively tested under high power pulse conditions in the SPERT Test Reactor. The fuel assemblies are rectangular in a horizontal cross section, 315 inches by 2.74 inches, allowing for flat control blades to be inserted in the core grid arrangement. The active height of the core is approximately 24 inches. In the initial Buffalo AMF contract, a collaborative development agreement was signed in conjunction with agreement to construct the facility. After completion of the Buffalo PULSTAR Reactor, the PULSTAR fuel underwent an extensive test program which resulted in some minor changes in the basic design. In 1965, North Carolina State University contracted with AMF for the construction of a dual MW steady state (with ultimate capability of 5 MW and pulsing PULSTAR Research Reactor. Their fuel is identical to the Buffalo fuel except for having an enrichment of 4% U-235. This paper presented basic information about the characteristics and performance of the PULSTAR Research Reactor fuel. The following summarizes this information. The fuel is of

  1. PULSTAR fuel, low enrichment, long lifetime, economical, proven

    Energy Technology Data Exchange (ETDEWEB)

    Carter, Robert E; Leonard, Bobby E [Institute for Resource Management, Inc., Bethesda, MD (United States)

    1993-08-01

    In 1962, the Western New York Research Center, Inc., located at the State University of New York at Buffalo, decided they had a need for a reactor with pulsing and high power steady state capabilities. Both General Atomic and the American Machine and Foundry Corporation (AMF) were contacted to ascertain if it were feasible to construct a dual purpose reactor of this type. The General Atomic proposal indicated the feasibility but would not warrant a steady state power of 2 MW with ultimate capability of 5 MW. AMF did provide a conceptual design for such a dual reactor, call the PULSTAR, and sufficient design information to confirm that the operating specifications could be met. The PULSTAR fuel consisted of 6 enrichment UO{sub 2} sintered pellets in zircaloy tubes (pins) mounted in a x 5 array inside a fuel assembly. The fuel design was patterned after fuel that was under development for light water power reactors and that had been extensively tested under high power pulse conditions in the SPERT Test Reactor. The fuel assemblies are rectangular in a horizontal cross section, 315 inches by 2.74 inches, allowing for flat control blades to be inserted in the core grid arrangement. The active height of the core is approximately 24 inches. In the initial Buffalo AMF contract, a collaborative development agreement was signed in conjunction with agreement to construct the facility. After completion of the Buffalo PULSTAR Reactor, the PULSTAR fuel underwent an extensive test program which resulted in some minor changes in the basic design. In 1965, North Carolina State University contracted with AMF for the construction of a dual MW steady state (with ultimate capability of 5 MW and pulsing PULSTAR Research Reactor. Their fuel is identical to the Buffalo fuel except for having an enrichment of 4% U-235. This paper presented basic information about the characteristics and performance of the PULSTAR Research Reactor fuel. The following summarizes this information. The fuel

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

    OpenAIRE

    Pešić Milan P.; Šotić Obrad; Hopwood William H.Jr

    2002-01-01

    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.

  3. 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.

  4. Low enrichment fuel conversion for Iowa State University. Final report

    International Nuclear Information System (INIS)

    Bullen, D.B.; Wendt, S.E.

    1996-01-01

    The UTR-10 research and teaching reactor at Iowa State University (ISU) has been converted from high-enriched fuel (HEU) to low- enriched fuel (LEU) under Grant No. DE-FG702-87ER75360 from the Department of Energy (DOE). The original contract period was August 1, 1987 to July 31, 1989. The contract was extended to February 28, 1991 without additional funding. Because of delays in receiving the LEU fuel and the requirement for disassembly of the HEU assemblies, the contract was renewed first through May 31, 1992, then through May 31, 1993 with additional funding, and then again through July 31, 1994 with no additional funding. In mid-August the BMI cask was delivered to Iowa State. Preparations are underway to ship the HEU fuel when NRC license amendments for the cask are approved

  5. A simplified treatment of radial enrichment distributions of LWR fuel assemblies in criticality calculations

    International Nuclear Information System (INIS)

    Hennebach, M.; Schnorrenberg, N.

    2008-01-01

    Criticality safety assessments are usually performed for fuel assembly models that are as generic as possible to encompass small modifications in geometry that have no impact on criticality. Dealing with different radial enrichment distributions for a fuel assembly type, which is especially important for BWR fuel, poses more of a challenge, since this characteristic is rather obviously influencing the neutronic behaviour of the system. Nevertheless, the large variability of enrichment distributions makes it very desirable and even necessary to treat them in a generalized way, both to keep the criticality safety assessment from becoming too unwieldy and to avoid having to extend it every time a new variation comes up. To be viable, such a generic treatment has to be demonstrably covering, i.e. lead to a higher effective neutron multiplication factor k eff than any of the radial enrichment distributions it represents. Averaging the enrichment evenly over the fuel rods of the assembly is a general and simple approach, and under reactor conditions, it is also a covering assumption: the graded distribution is introduced to achieve a linear power distribution, therefore reducing the enrichment of the better moderated rods at the edge of the assembly. With an even distribution of the average enrichment over all rods, these wellmoderated rods will cause increased fission rates at the assembly edges and a rise in k eff . Since the moderator conditions in a spent nuclear fuel cask differ strongly from a reactor even when considering optimal moderation, the proof that a uniform enrichment distribution is a covering assumption compared with detailed enrichment distributions has to be cask-specific. In this report, a method for making that proof is presented along with results for fuel assemblies from BWR reactors. All results are from three-dimensional Monte Carlo calculations with the SCALE 5.1 code package [1], using a 44-group neutron crosssection library based on ENDF

  6. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F.

    2015-01-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW e and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k eff = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  7. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

    2015-07-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  8. A nondestructive testing device for determining 235U enrichment in power reactor fuel elements

    International Nuclear Information System (INIS)

    Liu Lanhua; Liu Nangai

    1990-07-01

    The development and application of a nondestructive testing device are presented, which is used for determining the 235 U enrichment in the mixed fuel of fuel elements with UO 2 pellets. The testing efficiency is improved because the passive gamma ray method and a hole-bored NaI crystal and four channel multichannel analyzer are used. The false discrimination rate is reduced as the average comparing method is taken. This device is simple in structure and easy in operation. It has provided a new testing tool for the fuel elements production in China. This device has successfully been used in Qinshan Nuclear Power Plant in testing its fuel elements

  9. The U.S. reduced enrichment research and test reactor (RERTR) program

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-01-01

    Research and test reactors are widely deployed to study the irradiation behavior of materials of interest in nuclear engineering, to produce radioisotopes for medicine, industry, and agriculture, and as a basic research and teaching tool. In order to maximize neutron flux per unit power and/or to minimize capital costs and fuel cycle costs, most of these reactors were de- signed to utilize uranium with very high enrichment (in the 70% to 95% range). Research reactor fuels with such high uranium enrichment cause a potential risk of nuclear weapons proliferation. Over 140 research and test reactors of significant power (between 10 kW and 250 MW) are in operation with very highly enriched uranium in more than 35 countries, with total power in excess of 1,700 MW. The overall annual fuel requirement of these reactors corresponds to approximately 1,200 kg of 235 U. This highly strategic material is normally exported from the United States, converted to metal form, shipped to a fuel fabricator, and then shipped to the reactor site in finished fuel element form. At the reactor site the fuel is first stored, then irradiated, stored again, and eventually shipped back to the United States for reprocessing. The whole cycle takes approximately four years to complete, bringing the total required fuel inventory to approximately 5,000 kg of 235 U. The resulting international trade in highly-enriched uranium may involve several countries in the process of refueling a single reactor and creates a considerable concern that the highly-enriched uranium may be diverted for non-peaceful purposes while in fabrication, transport, or storage, particularly when it is in the unirradiated form. The proliferation resistance of nuclear fuels used in research and test reactors can be considerably improved by reducing their uranium enrichment to a value less than 20%, but significantly greater than natural to avoid excessive plutonium production

  10. Configuration of LWR fuel enrichment or burnup yielding maximum power

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

    An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

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

    International Nuclear Information System (INIS)

    Rydell, R.J.

    1990-01-01

    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)

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

    International Nuclear Information System (INIS)

    Rojas C, E.L.

    1998-01-01

    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)

  13. Linear accelerator fuel enricher regenerator (LAFER) and fission product transmutor (APEX)

    International Nuclear Information System (INIS)

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

    1979-01-01

    In addition to safety, two other major problems face the nuclear industry today; first is the long-term supply of fissle material and second is the disposal of long-lived fission product waste. The higher energy proton linear accelerator can assist in the solution of each of these problems. High energy protons from the linear accelerator interact with a molten lead target to produce spallation and evaporation neutrons. The neutrons are absorbed in a surrounding blanket of light water power reactor (LWR) fuel elements to produce fissile Pu-239 or U-233 fuel from natural fertile U-238 or Th-232 contained in the elements. The fissile enriched fuel element is used in the LWR power reactor until its reactivity is reduced after which the element is regenerated in the linear accelerator target/blanket assembly and then the element is once again burned (fissioned) in the power LWR. In this manner the natural uranium fuel resource can supply an expanding nuclear power reactor economy without the need for fuel reprocessing, thus satisfying the US policy of non-proliferation. In addition, the quantity of spent fuel elements for long-term disposal is reduced in proportion to the number of fuel regeneration cycles through the accelerator. The limiting factor for in-situ regeneration is the burnup damage to the fuel cladding material. A 300 ma-1.5 GeV (450 MW) proton linear accelerator can produce approximately one ton of fissile (Pu-239) material annually which is enough to supply fuel to three 1000 MW(e) LWR power reactors. With two cycles of enriching and regenerating, the nuclear fuel natural resource can be stretched by a factor of 3.6 compared to present fuel cycle practice without the need for reprocessing. Furthermore, the need for isotopic enrichment facilities is drastically reduced

  14. The proposed use of low enriched uranium fuel in the High Flux Australian Reactor (HIFAR)

    International Nuclear Information System (INIS)

    Vittorio, D.; Durance, G.

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO) operates the High Flux Australian Reactor (HIFAR). HIFAR commenced operation in the late 1950's with fuel elements containing uranium enriched to 93%. From that time the level of enrichment has gradually decreased to the current level of 60%. It is now proposed to further reduce the enrichment of HIFAR fuel to <20% by utilising LEU fuel assemblies manufactured by RISO National Laboratory, that were originally intended for use in the DR-3 reactor. Minor modifications have been made to the assemblies to adapt them for use in HIFAR. A detailed design review has been performed and initial safety analysis and reactor physics calculations are to be submitted to ARPANSA as part of a four-stage approval process. (author)

  15. The burnable poisons utilization for fissile enriched CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)

    1996-12-31

    Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 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. The current state of the Russian reduced enrichment research reactors program

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A. [and others

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

  18. Status of the reduced enrichment for research reactors program in Argentina

    International Nuclear Information System (INIS)

    Perez, E.; Kohut, C.

    2004-01-01

    In the area of Research and Test Reactors' fuel elements, the different stages of development carried out by the Atomic Energy Commission of Argentina (CNEA) until now, and the future plans are presented in this paper. Own and foreign programs, for reducing the risk of proliferation due the use of high enriched uranium fuel elements in these types of reactors, is mentioned. A brief description of different work performed is presented: At first the experience with the use of highly enriched uranium, and then the activities related with the development done in order to achieve a good knowledge in low-enriched (LEU) fuels, particularly in the area of U308-Al fuels. This experience has permitted us, supported by the excellent results obtained, to be in a position to satisfy our own requirements and also to supply to other countries, not only fuels but also technology transferences and facilities of the development appropriate for this purpose. The main modifications brought in the design and fabrication of these types of fuel elements is also described. Finally, and with the main objective to complete the development and to qualify the LEU fuels based on silicides and to improve the actual MO-99 blanket fabrication technology two new C.N.E.A. projects, are outlined.(author)

  19. Low-enriched uranium-molybdenum fuel plate development

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. Choice and utilization of slightly enriched uranium fuel for high performance research reactors

    International Nuclear Information System (INIS)

    Cerles, J.M.; Schwartz, J.P.

    1978-01-01

    Problems relating to the replacement of highly enriched (90% or 93% U 235 ) uranium fuel: by moderately enriched (20% or 40% in U 235 ) metallic uranium fuel and slightly enriched (3% or 8% in U 235 ) uranium oxide fuel are discussed

  1. Development of very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

  2. Use of enriched uranium as a fuel in CANDU reactors

    International Nuclear Information System (INIS)

    Zech, H.J.

    1976-08-01

    The use of slightly enriched uranium as a fuel in CANDU-reactors is studied in a simple parametric way. The results show the possibility of 1) about 30% savings in natural uranium consumption 2) about 35% increase in the utilization of the natural uranium 3) a decrease in fuelling costs to about 70 - 80% of the normal case of natural uranium fuelling. (orig.) [de

  3. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

    1987-01-01

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  4. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  5. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately

  6. Low enriched uranium fuel conversion and fuel shipping guide

    International Nuclear Information System (INIS)

    1997-01-01

    The analysis of reactor core physics and thermal hydraulics was completed in 1993. A supplement to the Final Safety Analysis Report describing the results of these analyses was submitted to the Nuclear Regulatory Commission along with proposed Technical Specifications in May, 1993. Discussions with the NRC staff led to a submittal of revised proposed Technical Specifications in February, 1994. The analytical work is complete. A second portion of the grant was to develop a fuel shipping guide for university research reactors. Such a guide was developed and is available for use by the research reactor community

  7. An enhanced search algorithm for the charged fuel enrichment in equilibrium cycle analysis of REBUS-3

    International Nuclear Information System (INIS)

    Park, Tongkyu; Yang, Won Sik; Kim, Sang-Ji

    2017-01-01

    Highlights: • An enhanced search algorithm for charged fuel enrichment was developed for equilibrium cycle analysis with REBUS-3. • The new search algorithm is not sensitive to the user-specified initial guesses. • The new algorithm reduces the computational time by a factor of 2–3. - Abstract: This paper presents an enhanced search algorithm for the charged fuel enrichment in equilibrium cycle analysis of REBUS-3. The current enrichment search algorithm of REBUS-3 takes a large number of iterations to yield a converged solution or even terminates without a converged solution when the user-specified initial guesses are far from the solution. To resolve the convergence problem and to reduce the computational time, an enhanced search algorithm was developed. The enhanced algorithm is based on the idea of minimizing the number of enrichment estimates by allowing drastic enrichment changes and by optimizing the current search algorithm of REBUS-3. Three equilibrium cycle problems with recycling, without recycling and of high discharge burnup were defined and a series of sensitivity analyses were performed with a wide range of user-specified initial guesses. Test results showed that the enhanced search algorithm is able to produce a converged solution regardless of the initial guesses. In addition, it was able to reduce the number of flux calculations by a factor of 2.9, 1.8, and 1.7 for equilibrium cycle problems with recycling, without recycling, and of high discharge burnup, respectively, compared to the current search algorithm.

  8. Criticality safety study of dry spent fuel cask loaded with increased enrichment fuel

    International Nuclear Information System (INIS)

    Bznuni, S.; Baghdasaryan, N.; Amirjanyan, A.

    2013-01-01

    Existing Dry Spent Fuel Casks (DSC) for transporting and storing of Armenian NPP fuel was licensed for WWER-440 fuel assemblies with 3.6% enrichment. Having in mind that ANPP introduced new fuel assemblies with increased enrichment (3.82 %) re-assessment of criticality safety analysis for DSC is required. Criticality safety analysis of DSC was performed by KENO-VI program using 238-GROUP ENDF/B-VII.0 LIBRARY (V7-238). Results of analysis showed that additional 8 borated racks for fuel assemblies should be included in the design of DSC. In addition feasibility study was performed to find out level of burnup-credit approach implementation to keep current design of DSC unchanged. Burnup-credit analysis was performed by STARBUCS program using axial burnup profiles from Armenian NPP neutronics analysis carried out by BIPR code. (authors)

  9. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments

    International Nuclear Information System (INIS)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E.; Xolocostli M, J. V.

    2008-01-01

    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)

  10. Operational impacts of low-enrichment uranium fuel conversion on the Ford Nuclear Reactor

    International Nuclear Information System (INIS)

    Bernal, F.E.; Brannon, C.C.; Burgard, N.E.; Burn, R.R.; Cook, G.M.; Simpson, P.A.

    1985-01-01

    The University of Michigan Department of Nuclear Engineering and the Michigan Memorial-Phoenix Project have been engaged in a cooperative effort with Argonne National Laboratory to test and analyze low-enrichment fuel in the Ford Nuclear Reactor (FNR). The effort was begun in 1979, as part of the Reduced Enrichment Research and Test Reactor Program, to demonstrate on a whole-core basis the feasibility of enrichment reduction from 93% to <20% in Materials Test Reactor-type fuel designs. The first low-enrichment uranium (LEU) core was loaded into the FNR and criticality was achieved on December 8, 1981. The final LEU core was established October 11, 1984. No significant operational impacts have resulted from conversion of the FNR to LEU fuel. Thermal flux in the core has decreased slightly; thermal leakage flux has increased. Rod worths, temperature coefficient, and void coefficient have changed imperceptibly. Impressions from the operators are that power defect has increased slightly and that fuel lifetime has increased

  11. Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process

    International Nuclear Information System (INIS)

    Heckendorn, F.M.

    2001-01-01

    Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-01

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

  13. Study on Characteristics of Co-firing Ammonia/Methane Fuels under Oxygen Enriched Combustion Conditions

    Science.gov (United States)

    Xiao, Hua; Wang, Zhaolin; Valera-Medina, Agustin; Bowen, Philip J.

    2018-06-01

    Having a background of utilising ammonia as an alternative fuel for power generation, exploring the feasibility of co-firing ammonia with methane is proposed to use ammonia to substitute conventional natural gas. However, improvement of the combustion of such fuels can be achieved using conditions that enable an increase of oxygenation, thus fomenting the combustion process of a slower reactive molecule as ammonia. Therefore, the present study looks at oxygen enriched combustion technologies, a proposed concept to improve the performance of ammonia/methane combustion. To investigate the characteristics of ammonia/methane combustion under oxygen enriched conditions, adiabatic burning velocity and burner stabilized laminar flame emissions were studied. Simulation results show that the oxygen enriched method can help to significantly enhance the propagation of ammonia/methane combustion without changing the emission level, which would be quite promising for the design of systems using this fuel for practical applications. Furthermore, to produce low computational-cost flame chemistry for detailed numerical analyses for future combustion studies, three reduced combustion mechanisms of the well-known Konnov's mechanism were compared in ammonia/methane flame simulations under practical gas turbine combustor conditions. Results show that the reduced reaction mechanisms can provide good results for further analyses of oxygen enriched combustion of ammonia/methane. The results obtained in this study also allow gas turbine designers and modellers to choose the most suitable mechanism for further combustion studies and development.

  14. Airborne effluent control at fuel enrichment, conversion, and fabrication plants

    International Nuclear Information System (INIS)

    Mitchell, M.E.

    1976-01-01

    Uranium conversion, enrichment, and fuel fabrication facilities generate gaseous wastes that must be treated prior to being discharged to the atmosphere. Since all three process and/or handle similar compounds, they also encounter similar gaseous waste disposal problems, the majority of which are treated in a similar manner. Ventilation exhausts from personnel areas and equipment off-gases that do not contain corrosive gases (such as HF) are usually passed through roughening and/or HEPA filters prior to release. Ventilation exhausts that contain larger quantities of particles, such as the conversion facilities' U 3 O 8 sampling operation, are passed through bag filters or cyclone separators, while process off-gases containing corrosive materials are normally treated by sintered metal filters or scrubbers. The effectiveness of particle removal varies from about 90 percent for a scrubber alone to more than 99.9 percent for HEPA filters or a combination of the various filters and scrubbers. The removal of nitrogen compounds (N 2 , HNO 3 , NO/sub x/, and NH 3 ) is accomplished by scrubbers in the enrichment and fuel fabrication facilities. The conversion facility utilizes a nitric acid recovery facility for both pollution control and economic recovery of raw materials. Hydrogen removal from gaseous waste streams is generally achieved with burners. Three different systems are currently utilized by the conversion, enrichment, and fuel fabrication plants to remove gaseous fluorides from airborne effluents. The HF-rich streams, such as those emanating from the hydrofluorination and fluorine production operations of the conversion plant, are passed through condensers to recover aqueous hydrofluoric acid

  15. Optimization of axial enrichment distribution for BWR fuels using scoping libraries and block coordinate descent method

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2017-03-15

    Highlights: • An optimization method for axial enrichment distribution in a BWR fuel was developed. • Block coordinate descent method is employed to search for optimal solution. • Scoping libraries are used to reduce computational effort. • Optimization search space consists of enrichment difference parameters. • Capability of the method to find optimal solution is demonstrated. - Abstract: An optimization method has been developed to search for the optimal axial enrichment distribution in a fuel assembly for a boiling water reactor core. The optimization method features: (1) employing the block coordinate descent method to find the optimal solution in the space of enrichment difference parameters, (2) using scoping libraries to reduce the amount of CASMO-4 calculation, and (3) integrating a core critical constraint into the objective function that is used to quantify the quality of an axial enrichment design. The objective function consists of the weighted sum of core parameters such as shutdown margin and critical power ratio. The core parameters are evaluated by using SIMULATE-3, and the cross section data required for the SIMULATE-3 calculation are generated by using CASMO-4 and scoping libraries. The application of the method to a 4-segment fuel design (with the highest allowable segment enrichment relaxed to 5%) demonstrated that the method can obtain an axial enrichment design with improved thermal limit ratios and objective function value while satisfying the core design constraints and core critical requirement through the use of an objective function. The use of scoping libraries effectively reduced the number of CASMO-4 calculation, from 85 to 24, in the 4-segment optimization case. An exhausted search was performed to examine the capability of the method in finding the optimal solution for a 4-segment fuel design. The results show that the method found a solution very close to the optimum obtained by the exhausted search. The number of

  16. The low enriched uranium fuel cycle in Ontario

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-02-01

    Six fuel-cycle strategies for use in CANDU reactors are examined in terms of their uranium-conserving properties and their ease of commercialization for three assumed growth rates of installed nuclear capacity in Ontario. The fuel cycle strategies considered assume the continued use of the natural uranium cycle up to the mid-1990's. At that time, the low-enriched uranium (LEU) cycle is gradually introduced into the existing power generation grid. In the mid-2020's one of four advanced cycles is introduced. The advanced cycles considered are: mixed oxide, intermediate burn-up thorium (Pu topping), intermediate burn-up thorium (U topping), and LMFBR. For comparison purposes an all natural uranium strategy and a natural uranium-LEU strategy (with no advanced cycle) are also included. None of the strategies emerges as a clear, overall best choice. (LL)

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

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    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 (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -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 U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -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 U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  18. Neutronic analysis of a fuel element with variations in fuel enrichment and burnable poison

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Martins, Felipe; Velasquez, Carlos E.; Cardoso, Fabiano; Fortini, Angela; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    In this work, the goal was to evaluate the neutronic behavior during the fuel burnup changing the amount of burnable poison and fuel enrichment. For these analyses, it was used a 17 x 17 PWR fuel element, simulated using the 238 groups library cross-section collapsed from ENDF/BVII.0 and TRITON module of SCALE 6.0 code system. The results confirmed the effective action of the burnable poison in the criticality control, especially at Beginning Of Cycle (BOC) and in the burnup kinetics, because at the end of the fuel cycle there was a minimal residual amount of neutron absorbers ({sup 155}Gd and {sup 157}Gd), as expected. At the end of the cycle, the fuel element was still critical in all simulated situations, indicating the possibility of extending the fuel burn. (author)

  19. The RERTR [Reduced Enrichment Research and Test Reactor] Program: Progress and plans

    International Nuclear Information System (INIS)

    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 3 Si 2 -Al and U 3 Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 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 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 3 Si-Al with 19.75% enrichment and U 3 Si 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

  20. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels guidebook

    International Nuclear Information System (INIS)

    1980-08-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this document has been prepared to assist reactor operators in determining whether conversion to the use of low enriched uranium (LEU) fuel designs is technically feasible for their specific reactor, and to assist in making a smooth transition to the use of LEU fuel designs where appropriate

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    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 (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -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 U 3 O 8 -Al was about 2% more than the original UAl x -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.

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

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-09-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    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.

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

    International Nuclear Information System (INIS)

    2004-01-01

    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

  7. Progress in qualifying low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

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

    2001-01-01

    The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

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

    International Nuclear Information System (INIS)

    James, R.A.

    1980-01-01

    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)

  9. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico... Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized...

  10. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice... Louisiana Energy Services (LES), LLC, National enrichment Facility in Eunice, New Mexico, and has verified...

  11. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico..., National Enrichment Facility in Eunice, New Mexico, and has authorized the introduction of uranium...

  12. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice... Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has verified that cascades...

  13. Uranium-236 in light water reactor spent fuel recycled to an enriching plant

    International Nuclear Information System (INIS)

    de la Garza, A.

    1977-01-01

    The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

  14. Uranium Enrichment Determination of the InSTEC Sub Critical Ensemble Fuel by Gamma Spectrometry

    International Nuclear Information System (INIS)

    Borrell Munnoz, Jose L.; LopezPino, Neivy; Diaz Rizo, Oscar; D'Alessandro Rodriguez, Katia; Padilla Cabal, Fatima; Arbelo Penna, Yunieski; Garcia Rios, Aczel R.; Quintas Munn, Ernesto L.; Casanova Diaz, Amaya O.

    2009-01-01

    Low background gamma spectrometry was applied to analyze the uranium enrichment of the nuclear fuel used in the InSTEC Sub Critical ensemble. The enrichment was calculated by two variants: an absolute method using the Monte Carlo method to simulated detector volumetric efficiency, and an iterative procedure without using standard sources. The results confirm that the nuclear fuel of the ensemble is natural uranium without any additional degree of enrichment. (author)

  15. Low enrichment fuel conversion for Iowa State University

    International Nuclear Information System (INIS)

    Rohach, A.F.; Hendrickson, R.A.

    1990-08-01

    Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. The LEU fuel has not been received. The HEU fuel assemblies for the UTR-10 reactor will not fit into any current research reactor shipping containers; therefore, the fuel assemblies must be disassembled and the fuel shipped as fuel plates. Procedures and practices have been developed so that the fuel assemblies will be disassembled in a shielded environment

  16. Improved locations of reactivity devices in future CANDU reactors fuelled with natural uranium or enriched fuels

    International Nuclear Information System (INIS)

    Boczar, P.G.; Van Dyk, M.T.

    1987-02-01

    A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg

  17. Development of very-high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

  18. Criticality issues with highly enriched fuels in a repository environment

    International Nuclear Information System (INIS)

    Taylor, L.L.; Sanchez, L.C.; Rath, J.S.

    1998-03-01

    This paper presents preliminary analysis of a volcanic tuff repository containing a combination of low enrichment commercial spent nuclear fuels (SNF) and DOE-owned SNF packages. These SNFs were analyzed with respect to their criticality risks. Disposal of SNF packages containing significant fissile mass within a geologic repository must comply with current regulations relative to criticality safety during transportation and handling within operational facilities. However, once the repository is closed, the double contingency credits for criticality safety are subject to unremediable degradation, (e.g., water intrusion, continued presence of neutron absorbers in proximity to fissile material, and fissile material reconfiguration). The work presented in this paper focused on two attributes of criticality in a volcanic tuff repository for near-field and far-field scenarios: (1) scenario conditions necessary to have a criticality, and (2) consequences of a nuclear excursion that are components of risk. All criticality consequences are dependent upon eventual water intrusion into the repository and subsequent breach of the disposal package. Key criticality parameters necessary for a critical assembly are: (1) adequate thermal fissile mass, (2) adequate concentration of fissile material, (3) separation of neutron poison from fissile materials, and (4) sufficient neutron moderation (expressed in units of moderator to fissile atom ratios). Key results from this study indicated that the total energies released during a single excursion are minimal (comparable to those released in previous solution accidents), and the maximum frequency of occurrence is bounded by the saturation and temperature recycle times, thus resulting in small criticality risks

  19. Replacement of highly enriched uranium by medium or low-enriched uranium in fuels for research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    To exclude the possibility of an explosive use of the uranium obtained from an elementary chemical process, one needs to use a fuel less enriched than 20 weight percent in U 235 . This goal can be reached by two ways: 1. The low density fuels, i.e. U or U 3 O 8 /Al fuels. One has to increase their U content from 1.3 g U/cm 3 presently qualified under normal operation conditions. Several manufacturers such as CERCA in France developed these fuels with a near-term objective of about 2 g U/cm 3 and a long-term objective of 3 g U/cm 3 . 2. The high density fuels. They are the UO 2 Caramel plate type fuels now under consideration, and U 3 Si and UMo as a long-term potential

  20. Study of Fuel Rods Axial Enrichment Distribution Effect on the Neutronic Parameters of the Reactor Core

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Nasiri, S. H.

    2012-01-01

    Optimization of the fuel burn up is an important issue in nuclear reactor fuel management and technology. Radial enrichment distribution in the reactor core is a conventional method and axial enrichment is constant along the fuel rod. In this article, the effects of axial enrichment distribution variation on neutronic parameters of PWR core are studied. The axial length of the core is divided into ten sections, considering axial enrichment variation and leaving the existing radial enrichment distribution intact. This study shows that the radial and axial power peaking factors are decreased as compared with the typical conventional core. In addition, the first core lifetime lasts 30 days longer than normal PWR core. Moreover, at the same time boric acid density is 0.2 g/kg at the beginning of the cycle. The flux shape is also flat at the beginning of the cycle for the proposed configuration of the axially enrichment distribution.

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

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Sosa, M.A.

    1987-01-01

    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 UO 2 required. (Author)

  2. Determining method and device for enrichment distribution inside of fuel assembly

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi.

    1997-01-01

    An enrichment degree at an initial burning stage of each of fuel rods of a BWR type reactor assembly is divided into groups. The enrichment degree at the initial burning stage of each of the groups is inputted, and the burning period from the loading to the taking out is divided into a plurality of burning steps. Nuclear characteristics of fuel assemblies such as the power of fuel rods, R-factor and infinite multiplication factor in each of the burning steps are estimated. The enrichment degree of the group of enrichment degree at the initial burning stage and the estimated power of fuel rods in a reactor operation state during the burning step are stored in the memory. A sensitivity coefficient showing the amount of change of the power of fuel rods in the burning step relative to the change of the enrichment degree of the group of enrichment degree is evaluated. A weighing function in the burning step is inputted. The maximum value of the product of the weighing function and the power of fuel rods throughout the entire burning steps is determined as an aimed function. Optimization calculation is conducted for determining the enrichment degree of the group so as to minimize the aimed function thereby determining the distribution of the enrichment degree. (N.H.)

  3. Research reactors. Problems of fuel element enrichment reduction. Deliberations and comments

    International Nuclear Information System (INIS)

    1978-10-01

    This paper summarises the main data from the major research reactors in the Federal Republic of Germany utilising highly enriched uranium (HEU) and presently available fuel technology for their fuel elements. The required modification for an adaption of the fabrication to lower enriched fuel are considered as well as the consequences on reactor performance operation and licensing. On the basis of past experience with reactor modifications a rough estimate of 82 months is given for the conversion of a reactor to a modified type of fuel and of 70 months for a fuel test program. The conclusions reflect the own calculations and data from other papers submitted to INFCE-WG 8C

  4. Contemporary and prospective fuel cycles for WWER-440 based on new assemblies with higher uranium capacity and higher average fuel enrichment

    International Nuclear Information System (INIS)

    Gagarinskiy, A.A.; Saprykin, V.V.

    2009-01-01

    RRC 'Kurchatov Institute' has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for WWER-440 reactors. Works were performed to upgrade and improve WWER-440 fuel cycles on the basis of second-generation fuel assemblies allowing core thermal power to be uprated to 107 108 % of its nominal value (1375 MW), while maintaining the same fuel operation lifetime. Currently intensive work is underway to develop fuel cycles based on second-generation assemblies with higher fuel capacity and average fuel enrichment per assembly increased up to 4.87 % of U-235. Fuel capacity of second-generation assemblies was increased by means of eliminated central apertures of fuel pellets, and pellet diameter extended due to reduced fuel cladding thickness. This paper intends to summarize the results of works performed in the field of WWER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of WWER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (Authors)

  5. Reduced enrichment for research and test reactors: proceedings

    International Nuclear Information System (INIS)

    1985-07-01

    Separate abstracts are presented for each of the papers included in the data base concerning RERTR programs and licensing; fuel development; plate-type fuel fabrication; fuel demonstration; economics; mixed cores; and applications

  6. Neutronics substantiation of possibility for conversion of the WWR-K reactor core to operation with low-enriched fuel

    International Nuclear Information System (INIS)

    Arinkin, F.M.; Gizatulin, Sh.H.; Zhantikin, T.M.; Koltochnik, S.N.; Takibaev, A.Zh.; Talanov, S.V.; Chakrov, P.V.; Chekushina, L.V.

    2002-01-01

    The studies are aimed to calculation and experimental justification of possibility for conversion of the WWR-R reactor core to low-enriched nuclear fuel (the 19.75-% enrichment in isotope U-235), resulting in reducing the risk of non-sanctioned proliferation of nuclear materials which can be used as weapons materials. The analysis of available published data, related to problem of reduction of enrichment in the fuel used in research thermal reactors, has been carried out. Basing on the analysis results, reference fuel compositions have been chosen, in particular, uranium dioxide (UO 2 ) in aluminum master form and the UA1 4 alloy. Preliminary calculations have shown that, with the WWR-K reactor core preserved existing critical characteristics (the fuel composition: UA1 4 ), the uranium concentration in the fuel element is to be increased by a factor of 2.0-2.2, being impossible technologically. The calculations have been performed by means of the Monte Carlo computational codes. The program of optimal conversion of the WWR-K reactor core to low-enriched fuel has been developed, including: development of calculation models of the reactor core, composed of various designs of fuel elements and fuel assemblies (FA), on a base of corresponding computational codes (diffusion, statistical, etc.); implementation of experiments in the zero-power reactor (critical assembly) with the WWR-C-type FA, in view of correction of the computational constants used in calculations; implementation of reactor core neutronics calculations, in view of selection of the U-235 optimal content in the low-enriched fuel elements and choice of FA reload strategy at the regime of reactor core after burning; determination of the fuel element specification; determination of the critical and operational loads for the reactor core composed of rod/tubular fuel elements; calculation of the efficiency of the protection control system effectors, optimization of its composition, number and locations in the

  7. 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.

  8. Evaluation of fuel performance with different enrichment degrees for an experimental device

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Pino, Eddy S.; Gomes, Daniel S.; Abe, Alfredo Y.; Silva, Antonio Teixeira e

    2013-01-01

    Evaluation of fuel performance is conventionally carried out using specific codes developed to this aim. The obtained data are confirmed by experimental measurements performed using devices, which are located inside research reactors, projected to simulate reactor conditions under normal operation. Due to the limitations of the available reactor core length for irradiation in research reactors core, fuel rods used to obtain experimental data must present the same characteristics of the real fuel rod, but with a shorter length. Then, in order to compare the obtained results to the expected behavior of the real fuel rod, the experimental fuel rod should be designed with a free volume to fuel volume ratio very closed to the one of the full scale fuel rod. The aim of this paper is to evaluate some parameters and aspects related to the fuel rod behavior in a rod applied to the experimental irradiation device called Nuclear Fuel Irradiation Circuit (CAFE-Mod1) considering two fuel enrichment degrees: a typical commercial PWR enrichment and a value about 4 times higher. This evaluation is carried out by means of an adapted fuel performance code. Some of the parameter evaluated were fuel temperature and fission gas release as function of the fuel enrichment level. The results obtained in this paper were very similar to the ones previously obtained without consider similar free volume between the experimental and the full length fuel rod, regardless of low increases observed for the internal rod pressure and the amount of fission gas released. (author)

  9. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  10. Irradiation behavior of low-enriched U/sub 6/Fe-Al dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, G.L.; Domagala, R.F.; Copeland, G.L.

    1987-10-01

    An irradiation test of miniature fuel plates containing low-enriched (20% /sup 235/U)U/sub 6/Fe dispersed and clad in Al was performed. The postirradiation examination shows U/sub 6/Fe to form extensive fission gas bubbles at burnups of only approx. = 20% of the original 20% fuel enrichment. Plate failure by fission gas-driven pillowing occurred at approx. = 40% burnup. This places U/sub 6/FE at the lowest burnup capability among low enriched dispersion fuels that have been tested for use in research and test reactors

  11. Environmental enrichment reduces signs of boredom in caged mink.

    Directory of Open Access Journals (Sweden)

    Rebecca K Meagher

    thus be operationalized and assessed empirically in non-human animals. It can also be reduced by environmental enrichment.

  12. The RERTR [Reduced Enrichment Research and Test Reactor] program: A progress report

    International Nuclear Information System (INIS)

    Travelli, A.

    1986-11-01

    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 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U 3 Si 2 -Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product 99 Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program

  13. Solid oxide fuel cells fueled with reducible oxides

    Science.gov (United States)

    Chuang, Steven S.; Fan, Liang Shih

    2018-01-09

    A direct-electrochemical-oxidation fuel cell for generating electrical energy includes a cathode provided with an electrochemical-reduction catalyst that promotes formation of oxygen ions from an oxygen-containing source at the cathode, a solid-state reduced metal, a solid-state anode provided with an electrochemical-oxidation catalyst that promotes direct electrochemical oxidation of the solid-state reduced metal in the presence of the oxygen ions to produce electrical energy, and an electrolyte disposed to transmit the oxygen ions from the cathode to the solid-state anode. A method of operating a solid oxide fuel cell includes providing a direct-electrochemical-oxidation fuel cell comprising a solid-state reduced metal, oxidizing the solid-state reduced metal in the presence of oxygen ions through direct-electrochemical-oxidation to obtain a solid-state reducible metal oxide, and reducing the solid-state reducible metal oxide to obtain the solid-state reduced metal.

  14. Determination of enrichment of recycle uranium fuels for different burnup values

    International Nuclear Information System (INIS)

    Zabunoglu, Okan H.

    2008-01-01

    Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

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

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1995-01-01

    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 U 3 Si 2 -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. (author)

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

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1991-01-01

    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 U 3 Si 2 -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

  17. International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the Embalse CANDU reactor

    International Nuclear Information System (INIS)

    Rouben, B.; Chow, H.C.; Leung, L.K.H.; Inch, W.; Fink, J.; Moreno, C.

    2004-01-01

    In the last few years, Nucleoelectrica Argentina S.A. and Atomic Energy of Canada Limited have collaborated on a study of the technical feasibility of implementing Slightly Enriched Uranium (SEU) fuel in the Embalse CANDU reactor in Argentina. The successful conversion to SEU fuel of the other Argentine heavy-water reactor, Atucha 1, served as a good example. SEU presents an attractive incentive from the point of view of fuel utilization: if fuel enriched to 0.9% 235 U were used in Embalse instead of natural uranium, the average fuel discharge burnup would increase significantly (by a factor of about 2), with consequent reduction in fuel requirements, leading to lower fuel-cycle costs and a large reduction in spent-fuel volume per unit energy produced. Another advantage is the change in the axial power shape: with SEU fuel, the maximum bundle power in a channel decreases and shifts towards the coolant inlet end, consequently increasing the thermalhydraulics safety margin. Two SEU fuel carriers, the traditional 37-element bundle and the 43-element CANFLEX bundle, which has enhanced thermalhydraulic characteristics as well as lower peak linear element ratings, have been examined. The feasibility study gave the organizations an excellent opportunity to perform cooperatively a large number of analyses, e.g., in reactor physics, thermalhydraulics, fuel performance, and safety. A Draft Plan for a Demonstration Irradiation of SEU fuel in Embalse was prepared. Safety analyses have been performed for a number of hypothetical accidents, such as Large Loss of Coolant, Loss of Reactivity Control, and an off-normal condition corresponding to introducing 8 SEU bundles in a channel (instead of 2 or 4 bundles). There are concrete safety improvements which result from the reduced maximum bundle powers and their shift towards the inlet end of the fuel channel. Further improvements in safety margins would accrue with CANFLEX. In conclusion, the analyses identified no issues that

  18. An optimal sequence of the reload charge fuel enrichment to a reactor

    International Nuclear Information System (INIS)

    Sato, S.

    1975-01-01

    An optimal sequence of enrichment of the reload charge of a three regions PWR during its life has been determined by dynamic programming. The state of the reactor is specified by the burnup of the fuel in the three regions and their initial enrichments. Constraints were imposed on the power peaking factor, the maximum burnup, the length of each stage between refueling and the total life of the reactor. 'Central-scatter loading' was assumed at each reloading. The two group diffusion equations were solved by the modal method for the static calculations of the reactor. Otimization of enrichment of the reload charge was performed under several hypotheses on the variation of the costs of uranium, costs of enrichment and the plant factor during the reactor life. It was observed that the optimum enrichment of the reload fuel is influenced more by the cost of enrichment rather than plant factor or cost of uranium. (Author) [pt

  19. Structure, conduct, and sustainability of the international low-enriched fuel fabrication industry

    International Nuclear Information System (INIS)

    Rothwell, Geoffrey

    2008-01-01

    This paper examines the cost structures of fabricating Low-Enriched Uranium fuel (LEU, enriched to 5% enrichment) light water reactor fuels. The LEU industry is decades old, and (except for high entry cost, i.e., the cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added by industry incumbents at Nth-of-a-Kind cost to the maximum capacity allowed by the license. On the other hand, new entrants face higher First-of-a-Kind costs and high new-facility licensing costs, increasing the scale required for entry thus discouraging small scale entry by countries with only a few nuclear power plants. Therefore, the industry appears to be competitive with sustainable investment in fuel-cycle states, and structural barriers-to-entry increase its proliferation resistance. (author)

  20. Specificity in the licensing process of reduced enrichment in the Bulgarian research reactor

    International Nuclear Information System (INIS)

    Vitkova, Marietta; Gorinov, Ivan

    2005-01-01

    The presented paper considers some specific questions of the licensing process regarding the reconstruction of the Bulgarian IRT-2000 research reactor, which includes conversion to the low enriched fuel. This specificity has risen as a result of two facts. The design of the reactor reconstruction was made on the basis of the existing fresh 36% highly enriched fuel. But after finishing of the design process, this fresh highly enriched fuel was shipped back to Russia in the framework of the RERTR program. These facts have involved some changes in both - in the licensing and the design processes. Re-analysis of the neutronic and thermal-hydraulic calculations is required to be made on the base of the technical specifications of the new LEU fuel. To facilitate the licensing process the NRA has adopted regulatory acceptance criteria for approval of the reactor core design with LEU fuel. (author)

  1. Conversion of the University of Missouri-Rolla Reactor from high-enriched uranium to low-enriched uranium fuel

    International Nuclear Information System (INIS)

    Bolon, A.E.; Straka, M.; Freeman, D.W.

    1997-01-01

    The objectives of this project were to convert the UMR Reactor fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel and to ship the HEU fuel back to the Department of Energy Savannah River Site. The actual core conversion was completed in the summer of 1992. The HEU fuel was offloaded to an onsite storage pit where it remained until July, 1996. In July, 1996, the HEU fuel was shipped to the DOE Savannah River Site. The objectives of the project have been achieved. DOE provided the following funding for the project. Several papers were published regarding the conversion project and are listed in the Attachment. In retrospect, the conversion project required much more time and effort than originally thought. Several difficulties were encountered including the unavailability of a shipping cask for several years. The authors are grateful for the generous funding provided by DOE for this project but wish to point out that much of their efforts on the conversion project went unfunded

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    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

  3. Low enrichment fuel conversion for Iowa State University

    International Nuclear Information System (INIS)

    Rohach, A.F.; Hendrickson, R.A.

    1991-08-01

    Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. This included development of procedures and tools for the disassembly process. During the period we held many practice sessions applying these tools and practices to a dummy fuel assembly. The LEU fuel was received on April 10, 1991 and the reactor was shut down on May 3, 1991 for refueling. The twelve HEU fuel assemblies in the UTR-10 reactor core were removed and disassembled during the week of May 6--9, 1991. The disassembly process went smoothly with only a few minor problems. Also during this reporting period several experimental measurements and preventative maintenance tasks were accomplished. Finally procedures and practices have been developed for the new LEU fuel loading and critical experiments which are to be completed during the late summer of 1991

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Status report - expert knowledge of operators in fuel reprocessing plants, enrichment plants and fuel fabrication plants

    International Nuclear Information System (INIS)

    Preuss, W.; Kramer, J.; Wildberg, D.

    1987-01-01

    The necessary qualifications of the responsible personnel and the knowledge required by personnel otherwise employed in nuclear plants are among the requirements for licensing laid down in paragraph 7 of the German Atomic Energy Act. The formal regulations for nuclear power plants are not directly applicable to plants in the fuel cycle because of the differences in the technical processes and the plant and work organisation. The aim of the project was therefore to establish a possible need for regulations for the nuclear plants with respect to the qualification of the personnel, and to determine a starting point for the definition of the required qualifications. An extensive investigation was carried out in the Federal Republic of Germany into: the formal requirements for training; the plant and personnel organisation structures; the tasks carried out by the responsible and otherwise employed personnel; and the state of training. For this purpose plant owners and managers were interviewed and the literature and plant specific documentation (e.g. plant rules) were reviewed. On the basis of literature research, foreign practices were determined and used to make comparative evaluations. The status report is divided into three separate parts for the reprocessing, the uranium enrichment, and the manufacture of the fuel elements. On the basis of the situation for reprocessing plants (particularly that of the WAK) and fuel element manufacturing plants, the development of a common (not uniform) regulation for all the examined plants in the fuel cycle was recommended. The report gives concrete suggestions for the content of the regulations. (orig.) [de

  6. Study of correcting the effect of daughter age on determining 235U enrichment of fuel rods

    International Nuclear Information System (INIS)

    Deng Jingshan; Zhou Chengfang; Luo Minxuan; Liu Yun

    1997-01-01

    Gamma-ray passive technique is a very effective method to assay and determine 235 U enrichment of nuclear power plant fuel rods. There is a weakness in this passive method, i.e. only after the uranium isotope daughters of UO 2 pellets have reached to equilibrium with uranium parent, then the 235 U enrichment can be determined. This weakness greatly restricts the application of the method. A new two-peak and two-window technique is developed that can overcome the interference of uranium daughter decay in determining 235 U enrichment of nuclear fuel rods, and the results are very satisfactory. The new technique will play an important role in the gamma-ray passive technique for determining 235 U enrichment of fuel rods. This new technique also makes the gamma-ray passive method perfectly. (11 figs., 6 tabs.)

  7. Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

    International Nuclear Information System (INIS)

    Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

    2014-01-01

    Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    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.

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

    International Nuclear Information System (INIS)

    1997-08-01

    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

  10. Atomics International fuel fabrication facility and low enrichment program [contributed by T.A. Moss, AI

    International Nuclear Information System (INIS)

    Moss, T.A.

    1993-01-01

    The AI facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area. Under the low enrichment program the AI fuel facility will be modified to accommodate a separate low enrichment Al x production line and compacting line. This facility modification should be done by the end of the fiscal year. We anticipate producing fuel with an enrichment slightly less than 20% We anticipate powder being available for plate production shortly after the facility is completed. Atomics International is scheduled to conduct plate LEU verification work using fully enriched material in the June-July time period, at which time we will investigate what level of uranium loadings we can go to using the current process. It is anticipated that 55 volume percent uranium compound in our fuel form can be achieved

  11. Some Main Results of Commissioning of the Dalat Research Reactor with Low Enriched Fuel

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2014-01-01

    After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements. (author)

  12. Welcome address to the 26th international meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    Sokolov, Y.

    2005-01-01

    While the IAEA has been a vigorous supporter of the RERTR programme since its inception. RERTR and the related fresh and spent fuel return efforts have gained new momentum with the launching of the Global Threat Reduction Initiative (GTRI) by U.S. Energy Secretary Abraham here in Vienna on May 25, 2004. All of the activities to be be discussed are included within the framework of the GTRI. The international programmes to qualify high density, LEU, dispersion fuels based on U-Mo alloys have run into unexpected technical difficulties that will delay qualification. A number of the presentations address the problems that have been encountered. At the same time, it is encouraging that the international resolve to reduce and eventually eliminate HEU in international commerce appears to have strengthened. In the past year, fresh HEU at research reactors in different countries have been returned to the country of origin. In all these examples, the return of the fresh fuel was accompanied by plans for conversion of existing reactors or design of new reactors to use LEU, as well as for the repatriation of spent research reactor fuel. The IAEA, particularly the Department of Technical Cooperation and my Department of Nuclear Energy has played an important role in implementing these fresh fuel return activities. In addition, several of the reactor conversion projects will be carried out under the auspices of IAEA technical cooperation projects and with important involvement of the Department of Nuclear Energy. The IAEA has also supported the repatriation of spent fuel to the country of original enrichment. The U.S. spent fuel acceptance programme has been operating for more than eight years, and was originally scheduled to terminate in 2006. Important announcements concerning the extension of the U.S. programme are expected. At the same time, the IAEA has been working hard with the U.S. and Russia to initiate the Russian research reactor spent fuel return programme. We are

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1984-05-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1984-05-01

    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.

  15. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Directory of Open Access Journals (Sweden)

    Mohammed Saad Dwiddar

    2015-08-01

    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  16. Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

    International Nuclear Information System (INIS)

    Pesic, M. P.

    1997-01-01

    Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

  17. Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133

    International Nuclear Information System (INIS)

    Skoda, R.; Rataj, J.; Uhera, J.

    2008-01-01

    The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)

  18. Reducing Actinide Production Using Inert Matrix Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark [Colorado School of Mines, Golden, CO (United States)

    2017-08-23

    The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

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

    International Nuclear Information System (INIS)

    Caprioli, Sara

    2004-04-01

    not fully burnt in the preceding cycles. In fact, the shorter cycle lengths of the transition cycles indicate that some fuel bundles are only partially burnt. These bundles add an excess of reactivity at the beginning of cycle 21. The excess of reactivity with which cycle 21 is loaded allows for a higher cycle burnup. One of the most relevant results arising from the entire work is the fact that the fuel bundles optimization in terms of the internal peaking factor plays a central role. In fact, the achievement of a lower and flatter internal peaking factor is much more significant than the acquired deviations and adjustments in the bundle reactivity. As seen in the cycle simulations, fresh fuel bundles with a higher internal peaking factor than in the respective standard bundles increase the core peaking factors. At the assembly level, the differences between assemblies with a central highly enriched region and a peripheric low enriched region, and assemblies with reversed configuration are not significant. At the core level, the relative position of these assembly differently configurated could play a significant role. In fact, if the radial neutron leakage is to be reduced from the periphery of the core, the low enriched fuel bundle regions should be placed towards the periphery of the core. In this case, the multiplication factor would play an important role in the core economy. However, it is always profitable to have a low internal peaking factor. The fact that cycle 21 carries all the desired features is certainly a promising result. Nevertheless, further simulation should be performed until equilibrium is achieved, that is, until the cycle parameters converge. Besides, one could investigate different geometries. The results could be more pronounced if variations in the average enrichment level of the bundle were allowed. Finally, an accurate safety and risk analysis, and economical calculations for the fuel types and the cores should be performed

  20. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E I; Jordanov, T; Christoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1994-12-31

    The idea of conversion of highly enriched uranium (HEU) from warheads without mixing it with natural uranium as well as the utilization of plutonium as fuel component is discussed. A nuclear fuel which is a mixture of 4% {sup 235}U (HEU) as a fissile isotope and 96 % {sup 232}Th (ThO{sub 2}) as a non-fissile isotope in a mixed oxide with thorium fuel is proposed. It is assumed that plutonium can also be used in the proposed fuel in a mixture with {sup 235}U. The following advantages of the use of HEU in LWRs in mixed {sup 235}U - Th fuel are pointed out: (1) No generation of long-living plutonium and americium isotopes (in case of reprocessing the high level radioactive wastes will contain only fission fragments and uranium); (2) The high conversion ratio of Th extends the expected burnup by approximately 1/3 without higher initial enrichment (the same initial enrichment simplifies the problem for compensation of the excess reactivity in the beginning with burnable poison and boric acid); (3) The high conversion ratio of Th allows the fuel utilization with less initial enrichment (by approx. 1/3) for the same burnup; thus less excess reactivity has to be compensated after reloading; in case of fuel reprocessing all fissile materials ({sup 235}U + {sup 233}U) could be chemically extracted. Irrespectively to the optimistic expectations outlined, further work including data on optimal loading and reloading schemes, theoretical calculations of thermal properties of {sup 235}U + Th fuel rods, manufacturing of several test fuel assemblies and investigations of their operational behaviour in a reactor core is still needed. 1 fig., 7 refs.

  1. Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

    Science.gov (United States)

    Bhattacharya, Sumit

    High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

  2. Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

    International Nuclear Information System (INIS)

    Hida, Kazuki; Yoshioka, Ritsuo

    1992-01-01

    A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

  3. K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1998-08-01

    This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

  4. Optimization of PWR fuel assembly radial enrichment and burnable poison location based on adaptive simulated annealing

    International Nuclear Information System (INIS)

    Rogers, Timothy; Ragusa, Jean; Schultz, Stephen; St Clair, Robert

    2009-01-01

    The focus of this paper is to present a concurrent optimization scheme for the radial pin enrichment and burnable poison location in PWR fuel assemblies. The methodology is based on the Adaptive Simulated Annealing (ASA) technique, coupled with a neutron lattice physics code to update the cost function values. In this work, the variations in the pin U-235 enrichment are variables to be optimized radially, i.e., pin by pin. We consider the optimization of two categories of fuel assemblies, with and without Gadolinium burnable poison pins. When burnable poisons are present, both the radial distribution of enrichment and the poison locations are variables in the optimization process. Results for 15 x 15 PWR fuel assembly designs are provided.

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

    KAUST Repository

    Crumlin, Ethan J.; Mutoro, Eva; Liu, Zhi; Grass, Michael E.; Biegalski, Michael D.; Lee, Yueh-Lin; Morgan, Dane; Christen, Hans M.; Bluhm, Hendrik; Shao-Horn, Yang

    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.

  6. Prompt neutron decay constant for the Oak Ridge Research Reactor with 20 wt % 235U enriched fuel

    International Nuclear Information System (INIS)

    Ragan, G.E.; Mihalczo, J.T.

    1986-01-01

    This paper describes measurements of the prompt neutron decay constant at delayed criticality for the Oak Ridge Research Reactor (ORR) using 20 wt % 235 U enriched fuel and compares these measurements with similar measurements using 93.2 wt % 235 U enriched fuel. This reactor parameter is of interest because it affects the transient behavior of the reactor in prompt criticality accident situations. This experiment is part of a program to investigate the differences in the performance of research reactors fueled with highly enriched and low enriched uranium. The prompt neutron decay constants were obtained using noise analysis measurement techniques for a core with newly fabricated, unirradiated fuel elements

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1991-10-01

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

  8. Experiments of JRR-4 low-enriched-uranium-silicied fuel core

    International Nuclear Information System (INIS)

    Hirane, Nobuhiko; Ishikuro, Yasuhiro; Nagadomi, Hideki; Yokoo, Kenji; Horiguchi, Hironori; Nemoto, Takumi; Yamamoto, Kazuyoshi; Yagi, Masahiro; Arai, Nobuyoshi; Watanabe, Shukichi; Kashima, Yoichi

    2006-03-01

    JRR-4, a light-water-moderated and cooled, swimming pool type research reactor using high-enriched uranium plate-type fuels had been operated from 1965 to 1996. In order to convert to low-enriched-uranium-silicied fuels, modification work had been carried out for 2 years, from 1996 to 1998. After the modification, start-up experiments were carried out to obtain characteristics of the low-enriched-uranium-silicied fuel core. The measured excess reactivity, reactor shutdown margin and the maximum reactivity addition rate satisfied the nuclear limitation of the safety report for licensing. It was confirmed that conversion to low-enriched-uranium-silicied fuels was carried out properly. Besides, the necessary data for reactor operation were obtained, such as nuclear, thermal hydraulic and reactor control characteristics. This report describes the results of start-up experiments and burnup experiments. The first criticality of low-enriched-uranium-silicied core was achieved on 14th July 1998, and the operation for joint-use has been carried out since 6th October 1998. (author)

  9. Report of the Working Party on the conversion of HIFAR to low enrichment uranium fuel

    International Nuclear Information System (INIS)

    1986-06-01

    This report states the effect on research reactor operations and applications of international and national political decisions relating to fuel enrichment. Technical work done in Australia and overseas to establish parameters for conversion of research reactors from High Enrichment Uranium (HEU) to Low Enrichment Uranium (LEU) have been considered in developing a strategy for HIFAR. The requirements of the research groups, isotope production group and reactor operating staff have been considered. For HIFAR to continue to provide the required facilities in support of the national need, it is concluded these should be no reduction of neutron flux

  10. Reduced enrichment for research and test reactors. Proceedings

    International Nuclear Information System (INIS)

    Thamm, G.; Brandt, M.

    1991-01-01

    The 12th meeting was attended by 113 participants coming from 21 countries and from EURATOM and IAEA.42 reports were presented orally within 10 sessions dealing with 5 main topics: 1) programs(5); 2) fuels(12); 3) reactor conversions(17); 5) high performance neutron sources(4); 5) others(4). (HP)

  11. Principal Angle Enrichment Analysis (PAEA): Dimensionally Reduced Multivariate Gene Set Enrichment Analysis Tool.

    Science.gov (United States)

    Clark, Neil R; Szymkiewicz, Maciej; Wang, Zichen; Monteiro, Caroline D; Jones, Matthew R; Ma'ayan, Avi

    2015-11-01

    Gene set analysis of differential expression, which identifies collectively differentially expressed gene sets, has become an important tool for biology. The power of this approach lies in its reduction of the dimensionality of the statistical problem and its incorporation of biological interpretation by construction. Many approaches to gene set analysis have been proposed, but benchmarking their performance in the setting of real biological data is difficult due to the lack of a gold standard. In a previously published work we proposed a geometrical approach to differential expression which performed highly in benchmarking tests and compared well to the most popular methods of differential gene expression. As reported, this approach has a natural extension to gene set analysis which we call Principal Angle Enrichment Analysis (PAEA). PAEA employs dimensionality reduction and a multivariate approach for gene set enrichment analysis. However, the performance of this method has not been assessed nor its implementation as a web-based tool. Here we describe new benchmarking protocols for gene set analysis methods and find that PAEA performs highly. The PAEA method is implemented as a user-friendly web-based tool, which contains 70 gene set libraries and is freely available to the community.

  12. REDUCING GREENHOUSE EMISSIONS AND FUEL CONSUMPTION

    Directory of Open Access Journals (Sweden)

    Susan A. SHAHEEN, Ph.D.

    2007-01-01

    Fortunately, transportation technologies and strategies are emerging that can help to meet the climate challenge. These include automotive and fuel technologies, intelligent transportation systems (ITS, and mobility management strategies that can reduce the demand for private vehicles. While the climate change benefits of innovative engine and vehicle technologies are relatively well understood, there are fewer studies available on the energy and emission impacts of ITS and mobility management strategies. In the future, ITS and mobility management will likely play a greater role in reducing fuel consumption. Studies are often based on simulation models, scenario analysis, and limited deployment experience. Thus, more research is needed to quantify potential impacts. Of the nine ITS technologies examined, traffic signal control, electronic toll collection, bus rapid transit, and traveler information have been deployed more widely and demonstrated positive impacts (but often on a limited basis. Mobility management approaches that have established the greatest CO2 reduction potential, to date, include road pricing policies (congestion and cordon and carsharing (short-term auto access. Other approaches have also indicated CO2 reduction potential including: low-speed modes, integrated regional smart cards, park-and-ride facilities, parking cash out, smart growth, telecommuting, and carpooling.

  13. Plutonium-enriched thermal fuel production experience in Belgium

    International Nuclear Information System (INIS)

    LeBlanc, J.M.

    1983-01-01

    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

  14. Selection and use of a low enriched fuel in high performance research reactors

    International Nuclear Information System (INIS)

    Cerles, J.M.; Schwartz, J.P.

    1978-08-01

    A new nuclear fuel composition for research reactors (Osiris, Siloe) is studied using low enriched (E<20%) uranium oxide. Its utilization leads to modifications in the facilities of these experimental reactors: increase of primary coolant flow, modifications in failed element detection system, handling of materials and storage

  15. Metallurgical and reactor physics aspects of using low enrichment fuel in Safari-I

    International Nuclear Information System (INIS)

    1978-09-01

    The feasibility of using lower than 93% enriched fuel in the SAFARI-I research and materials testing reactor is reviewed. Metallurgical experiments show that, using standard U-Al alloy technology and keeping the 235 U loading per element constant without altering the fuel plate thickness, a maximum of 35 weight percent of uranium in the meat can be achieved. This corresponds to using a minimum enrichment of 40% 235 U in order to retain the same mass of 235 U in the core. Even then a loss of approximately 3,3% in reactivity is calculated, which is more than the 2,8% sup(deltak)/k which is normally allowed for burnup. Using current U-Al alloy fuel technology, and an enrichment of approximately 45% 235 U, no changes in core configuration or coolant requirements will be necessary. The use of 20% enriched uranium will require the development of a new fuel design and technology if drastic redesign and modification of the reactor and coolant circuits is to be avoided. Without such new technology, the redesign and modification of the reactor will cost upwards of 3 million dollars and take up to 5 years to complete, requiring a complete shutdown of the reactor for approximately 2 years

  16. Analysis of the production of U3O8 powder for low enrichment fuel plates

    International Nuclear Information System (INIS)

    Boero, N.L.; Celora, J.; Parodi, C.A.; Ponieman, G.; Kellner, M.; Marajofsky, A.

    1987-01-01

    Description is made of the processes used in the production of U 3 O 8 powder for low enrichment plates for fuel elements for Research Reactors. The analysis of the efficiency of each batch is foccused on the relationship between milling and sieving times and the morphology of the product in each production step. (Author)

  17. Analysis Of The Effect Of Fuel Enrichment Error On Neutronic Properties Of The RSG-GAS Core

    International Nuclear Information System (INIS)

    Saragih, Tukiran; Pinem, Surian

    2002-01-01

    The analysis of the fuel enrichment error effect on neutronic properties has been carried out. The fuel enrichment could be improperly done because of wrong fabrication. Therefore it is necessary to analyze the fuel enrichment error effect to determine how many percents the fuel enrichment maximum can be accepted in the core. The analysis was done by simulation method The RSG-GAS core was simulated with 5 standard fuels and 1 control element having wrong enrichment when inserted into the core. Fuel enrichment error was then simulated from 20%, 25% and 30% and the simulation was done using WIMSD/4 and Batan-2DIFF codes. The cross section of core material of the RSG-GAS was generated by WIMSD/4 code in 1-D, X-Y geometry and 10 energy neutron group. Two dimensions, diffusion calculation based on finite element method was done by using Batan-2DIFF code. Five fuel elements and one control element changed the enrichment was finally arranged as a new core of the RSG-Gas reactor. The neutronic properties can be seen from eigenvalues (k eff ) as well as from the kinetic properties based on moderator void reactivity coefficient. The calculated results showed that the error are still acceptable by k eff 1,097 even until 25% fuel enrichment but not more than 25,5%

  18. Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC

    International Nuclear Information System (INIS)

    Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.

    1985-05-01

    The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs

  19. An experimental study of a hydrogen-enriched ethanol fueled Wankel rotary engine at ultra lean and full load conditions

    International Nuclear Information System (INIS)

    Amrouche, F.; Erickson, P.A.; Varnhagen, S.; Park, J.W.

    2016-01-01

    Highlights: • H_2 was added at the intake of a single-rotor ethanol fueled Wankel engine. • The engine was operating at ultra-lean condition, WOT and 3000 rpm. • H_2 enrichment helps shortening the burn duration, enhance the thermal efficiency and reduce the BSEC. • H_2 addition helps to reduce HC, CO and CO_2 emissions. - Abstract: In this paper, the effect of hydrogen addition to ethanol in a monorotor Wankel engine at wide open throttle position and in an ultra-lean operating regime was experimentally investigated. For this aim, variation of hydrogen enrichment levels on the ethanol engine performance and emissions were considered. Experiments were carried out under a constant engine speed of 3000 rpm and fixed spark timing of 15 °BTDC. The test results showed that hydrogen enrichment improved the combustion process through shortening of the flame development and flame propagation periods and reducing the cyclic variation. Furthermore, the reduction of burn duration with the increase of hydrogen fraction enhances the thermal efficiency, reducing the brake-specific energy consumption, as well as reducing the unburned hydrocarbons emissions of the Wankel engine.

  20. Evaluation of Biodiesel Fuels to Reduce Fossil Fuel Use in Corps of Engineers Floating Plant Operations

    Science.gov (United States)

    2016-07-01

    ER D C/ CH L TR -1 6- 11 Dredging Operations and Environmental Research Program Evaluation of Biodiesel Fuels to Reduce Fossil Fuel Use... Fuels to Reduce Fossil Fuel Use in Corps of Engineers Floating Plant Operations Michael Tubman and Timothy Welp Coastal and Hydraulics Laboratory...sensitive emissions, increase use of renewable energy, and reduce the use of fossil fuels was conducted with funding from the U.S. Army Corps of

  1. Development of IAEA safeguards at low enrichment uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Badawy, I.

    1988-01-01

    In this report the nuclear material at low enrichment uranium fuel fabrication plants under IAEA safeguards is studied. The current verification practices of the nuclear material and future improvements are also considered. The problems met during the implementation of the the verification measures of the nuclear material - particularly for the fuel assemblies are discussed. The additional verification activities as proposed for future improvements are also discussed including the physical inventory verification and the verification of receipts and shipments. It is concluded that the future development of the present IAEA verification practices at low enrichment uranium fuel fabrication plants would necessitate the application of quantitative measures of the nuclear material and the implementation of advanced measurement techniques and instruments. 2 fig., 4 tab

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

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Watanabe, Shouichi; Yoshioka, Kenichi; Mitsuhashi, Ishi; Kumanomido, Hironori; Sugahara, Satoshi; Hiraiwa, Kouji

    2009-01-01

    A concept of 'minimal-content gadolinia' with a content of less than several hundred ppm mixed in the 'above-5wt% enrichment UO 2 fuel' for super high burnup is proposed for ensuring the criticality safety in the UO 2 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 UO 2 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)

  3. U.S. progress in the development of very high density low enrichment research reactor fuels

    International Nuclear Information System (INIS)

    Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

    2012-01-01

    The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

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

    International Nuclear Information System (INIS)

    Muellauer, J.

    1989-01-01

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

  5. Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

    Science.gov (United States)

    Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

    1982-01-01

    A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

  6. Reduced size fuel cell for portable applications

    Science.gov (United States)

    Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

    2004-01-01

    A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

  7. Status of the natural and enriched uranium market: the basic economical factor for the development of the fuel cycle

    International Nuclear Information System (INIS)

    Nochev, T.

    1999-01-01

    Status of the Natural and Enriched Uranium Market - the Basic. Economical Factor for the Development of the Fuel Cycle An overview of the status of the natural and enriched uranium market has been performed and it offers a possibility to estimate the changes and tendencies, the knowledge of which is needed in negotiations about the fresh fuel. The simplified financial analysis presented here demonstrates the economical profitability of the storage of the spent fuel making now the allocations for the future reprocessing

  8. Conversion and standardization of university reactor fuels using low-enrichment uranium - Options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    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. The U.S. Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the U.S. Department of Energy. (author)

  9. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    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. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab

  10. Neutronic performance of a fusion-fission hybrid reactor designed for fuel enrichment for LWRs

    International Nuclear Information System (INIS)

    Yapici, H.; Baltacioglu, E.

    1997-01-01

    In this study, the breeding performance of a fission hybrid reactor was analyzed to provide fissile fuel for Light Water Reactors (LWR) as an alternative to the current methods of gas diffusion and gas centrifuge. LWR fuel rods containing UO 2 or ThO 2 fertile material were located in the fuel zone of the blanket and helium gas or Flibe (Li 2 BeF 4 ) fluid was used as coolant. As a result of the analysis, according to fusion driver (D,T and D,D) and the type of coolant the enrichment of 3%-4% were achieved for operation periods of 12 and 36 months in case of fuel rods containing UO 2 , respectively and for operation periods of 18 and 48 months in case of fuel rods containing ThO 2 , respectively. Depending on the type of fusion driver, coolant and fertile fuel, varying enrichments of between 3% and 8.9% were achieved during operation period of four years

  11. A disposition strategy for highly enriched, aluminum-based fuel from research and test reactors

    International Nuclear Information System (INIS)

    McKibben, J.M.; Gould, T.H.; McDonell, W.R.; Bickford, W.E.

    1994-01-01

    The strategy proposed in this paper offers the Department of Energy an approach for disposing of aluminum-based, highly enriched uranium (HEU) spent fuels from foreign and domestic research reactors. The proposal is technically, socially, and economically sound. If implemented, it would advance US non-proliferation goals while also disposing of the spent fuel's waste by timely and proven methods using existing technologies and facilities at SRS without prolonged and controversial storage of the spent fuel. The fuel would be processed through 221-H. The radioactive fission products (waste) would be treated along with existing SRS high level waste by vitrifying it as borosilicate glass in the Defense Waste Processing Facility (DWPF) for disposal in the national geological repository. The HEU would be isotopically diluted, during processing, to low-enriched uranium (LEU) which can not be used to make weapons, thus eliminating proliferation concerns. The LEU can be sold to fabricators of either research reactor fuel or commercial power fuel. This proposed processing-LEU recycle approach has several important advantages over other alternatives, including: Lowest capital investment; lowest net total cost; quickest route to acceptable waste form and final geologic disposal; and likely lowest safety, health, and environmental impacts

  12. Uranium enrichment

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.

    1988-06-01

    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

  13. Proceedings of the 1978 international meeting on reduced enrichment for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1993-08-01

    November 9-10, 1978, marked the first of what has become an annual event - the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The RERTR Program had been started only three months earlier, and the meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and while it established the basis for all later meetings, it was unique in several respects. It was a time of feeling each other out, and of sharing new ideas, concerns, and hopes. In the absence of an established precedent, a number of participants came with written papers while others made only verbal presentations. Informality added spice and special importance to the discussions at the end of each presentation and, especially, to the panel discussion at the end of the meeting. An important achievement was a consensus on near-, medium-, and long-term density goals for the various fuels. This consensus resulted in a list written on the blackboard at the end of the panel discussion, and reproduced on page 216, which outlined the goals of each fabricator. Luckily, both presentations and discussions were recorded on audio tape. These recordings were transcribed and used to complete the informal presentations and to append the discussions at the end of each presentation. Considerable effort was expended in clearing the transcribed papers and key discussions with the participants. A few issues could not be resolved quickly, and in the frantic rush of those early years, these proceedings were set aside. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort, we have recently dusted off the manuscript and finished our editing job.

  14. Proceedings of the 1978 international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-08-01

    November 9-10, 1978, marked the first of what has become an annual event - the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The RERTR Program had been started only three months earlier, and the meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and while it established the basis for all later meetings, it was unique in several respects. It was a time of feeling each other out, and of sharing new ideas, concerns, and hopes. In the absence of an established precedent, a number of participants came with written papers while others made only verbal presentations. Informality added spice and special importance to the discussions at the end of each presentation and, especially, to the panel discussion at the end of the meeting. An important achievement was a consensus on near-, medium-, and long-term density goals for the various fuels. This consensus resulted in a list written on the blackboard at the end of the panel discussion, and reproduced on page 216, which outlined the goals of each fabricator. Luckily, both presentations and discussions were recorded on audio tape. These recordings were transcribed and used to complete the informal presentations and to append the discussions at the end of each presentation. Considerable effort was expended in clearing the transcribed papers and key discussions with the participants. A few issues could not be resolved quickly, and in the frantic rush of those early years, these proceedings were set aside. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort, we have recently dusted off the manuscript and finished our editing job

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

    International Nuclear Information System (INIS)

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

    1988-07-01

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

  16. Alternative Fuels Data Center: Wisconsin Reduces Emissions With Natural Gas

    Science.gov (United States)

    Trucks Wisconsin Reduces Emissions With Natural Gas Trucks to someone by E-mail Share Alternative Fuels Data Center: Wisconsin Reduces Emissions With Natural Gas Trucks on Facebook Tweet about Alternative Fuels Data Center: Wisconsin Reduces Emissions With Natural Gas Trucks on Twitter Bookmark

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

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    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 UF 6 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

  18. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    The last decade worldwide experience of PWRs and WWERs confirms the trends for the improvement of the nuclear power industry electricity production through the implementation of high burn-up or high fuel duty, which are usually accompanied with the usage of UO 2 fuel with higher content of 235 U - 4.0% - 4.5% (5.0%). It was concluded that the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of the core are the primary and basic factors accompanying the implementation of uranium fuel with higher 235 U content, aiming extended fuel cycles and higher burn-up of the fuel in Pressurized Water Reactors. As main consequences of the presence of these factors the modifications of chemical / electrochemical environments of nuclear fuel cladding- and reactor coolant system- surfaces are evaluated. These conclusions are the reason for: 1) The determination of the choices of the type of fuel cladding materials in respect with their enough corrosion resistance to the specific fuel cladding environment, created by the presence of SNB; 2) The development and implementation of primary circuit water chemistry guidelines ensuring the necessary low corrosion rates of primary circuit materials and limitation of cladding deposition and out-of-core radioactivity buildup; 3) Implementation of additional neutron absorbers which allow enough decrease of the initial concentration of H 3 BO 3 in coolant, so that its neutralization will be possible with the permitted alkalising agent concentrations. In this paper the specific features of WWER-1000 units in Bulgarian Nuclear Power Plant; use of 235 U higher enriched fuel in the WWER-1000 reactors in the Kozloduy NPP; coolant water chemistry and radiochemistry plant data during the power operation period of the Kozloduy NPP Unit 5, 15 th fuel cycle; evaluation of the approaches and results by the conversion of the WWER-1000 Units at the Kozloduy NPP to the uranium fuel with 4.3% 235 U as

  19. Optimization of BWR fuel lattice enrichment and gadolinia distribution using genetic algorithms and knowledge

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Francois, Juan Luis; Carmona, Roberto; Oropeza, Ivonne P.

    2007-01-01

    An optimization methodology based on the Genetic Algorithms (GA) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The optimization algorithm was linked to the HELIOS code to evaluate the neutronic parameters included in the objective function. The goal is to search for a fuel lattice with the lowest average enrichment, which satisfy a reactivity target, a local power peaking factor (PPF), lower than a limit value, and an average gadolinia concentration target. The methodology was applied to the design of a 10 x 10 fuel lattice, which can be used in fuel assemblies currently used in the two BWRs operating at Mexico. The optimization process showed an excellent performance because it found forty lattice designs in which the worst one has a better neutronic performance than the reference lattice design. The main contribution of this study is the development of an efficient procedure for BWR fuel lattice design, using GA with an objective function (OF) which saves computing time because it does not require lattice burnup calculations

  20. Development of ISA procedure for uranium fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Arakawa, Tomoyuki; Yamashita, Masahiro; Sasaki, Noriaki; Hirano, Mitsumasa

    2011-01-01

    The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)

  1. Optimal management of fuel in nuclear reactors with slightly enriched uranium and heavy water

    International Nuclear Information System (INIS)

    Serghiuta, D.

    1994-01-01

    This Ph.D. thesis presents the general principles guiding the optimal management of the fuel in CANDU type reactors with slightly enriched uranium. A method is devised which is based on the specific physical characteristics of this type of reactors and makes use of the multipurpose mathematical programming satisfying economical and nuclear safety requirements. The main goal of this work was the establishing of a refueling optimal methodology at equilibrium maintaining the reactor critical during operation. It also minimizes the fuel cycle cost through minimization of the utilized fissile material and at the same time by maximizing the reactor duty time through an optimal chain of refilling operations. This work can be considered as a contribution to a future project of CANDU type reactor core based on slightly enriched uranium. 74 Figs., 9 Tabs., 62 Refs

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

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    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 UF 6 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. Nonproliferation and safeguards aspects of fuel cycle programs in reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1995-01-01

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO 2 fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO 2 once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments

  4. An approach to the nuclear fuel enrichment technology; Jedan prilaz tehnologiji obogacivanja nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Marsicanin, B [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1979-07-01

    In this paper the impact of new construction materials development on the technology of nuclear fuel enrichment by centrifugal method is considered. New composite materials, based on carbon fibres, with high tensile strength and low density have better characteristics than any other structural material used for centrifuge rotor so far. Possible improvements of centrifuge performance are pointed out, based on comparative analyses of material characteristics for composite and other materials. (author)

  5. Verification of the enrichment of fresh VVER-440 fuel assemblies at NPP Paks

    Energy Technology Data Exchange (ETDEWEB)

    Almasia, I.; Hlavathya, Z.; Nguyena, C. T. [Institute of Isotopes, Hungarian Academy of Sciences, Budapest, (Hungary); others, and

    2012-06-15

    A Non Destructive Analysis (NDA) method was developed for the verification of {sup 235}U enrichment of both homogeneous and profiled VVER-440 reactor fresh fuel assemblies by means of gamma spectrometry. A total of ca. 30 assemblies were tested, five of which were homogeneous, with {sup 235}U enrichment in the range 1,6% to 3,6%, while the others were profiled with pins of 3,3% to 4,4% enrichment. Two types of gamma detectors were used for the test measurements: 2 coaxial HPGe detectors and a miniature CdZnTe (CZT) detector fitting into the central tube of the assemblies. It was therefore possible to obtain information from both the inside and the outside of the assemblies. It was shown that it is possible to distinguish between different types of assemblies within a reasonable measurement time (about 1000 sec). For the HPGe measurements the assemblies had to be lifted out from their storage rack, while for the CZT detector measurements the assemblies could be left at their storage position, as it was shown that the neighbouring assemblies do not affect measurement inside the assemblies' central tube. The measured values were compared to Monte Carlo simulations carried out using the MCNP code, and a recommendation for the optimal approach to verify the {sup 235}U enrichment of fresh VVER-440 reactor fuel assemblies is suggested.

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

    International Nuclear Information System (INIS)

    Montierth, Leland M.

    2016-01-01

    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.

  7. 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.

  8. Improved performance of microbial fuel cells enriched with natural microbial inocula and treated by electrical current

    International Nuclear Information System (INIS)

    Lin, Hongjian; Wu, Xiao; Miller, Curtis; Zhu, Jun

    2013-01-01

    Microbial fuel cells (MFCs) are increasingly attracting attention as a sustainable technology as they convert chemical energy in organic wastes to electricity. In this study, the effects of different inoculum sources (river sediment, activated sludge and anaerobic sludge) and electrical current stimulation were evaluated using single-chamber air-cathode MFCs as model reactors based on performance in enrichment process and electrochemical characteristics of the reactors. The result revealed the rapid anodic biofilm development and substrate utilization of the anaerobic sludge-inoculated MFC. It was also found that the river sediment-inoculated MFC achieved the highest power output of 195 μW, or 98 mW m −2 , due to better developed anodic biofilm confirmed by scanning electron microscopy. The current stimulation enhanced the anodic biofilm attachment over time, and therefore reduced the MFC internal resistance by 27%, increased the electrical capacitance by four folds, and improved the anodic biofilm resilience against substrate deprivation. For mature MFCs, a transient application of a negative voltage (−3 V) improved the cathode activity and maximum power output by 37%. This improvement was due to the bactericidal effect of the electrode potential higher than +1.5 V vs. SHE, demonstrating a substantial benefit of treating MFC cathode after long-term operation using suitable direct electrical current. -- Highlights: •Voltage stimulation (+2 V) during inoculation reduced MFC internal resistance and improved biofilm resilience. •Voltage stimulation increased biofilm electrical capacitance by 5-fold. •Negative voltage stimulation (−3 V) enhanced the maximum power output by 37%. •River sediment MFC obtained higher power due to better anodic biofilm coverage. •Anaerobic sludge quickly developed anodic biofilm for MFC and quickly utilized volatile fatty acids

  9. 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

  10. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    Energy Technology Data Exchange (ETDEWEB)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

    2010-07-01

    Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.

  11. How can Korea secure uranium enrichment and spent fuel reprocessing rights?

    International Nuclear Information System (INIS)

    Roh, Seungkook; Kim, Wonjoon

    2014-01-01

    South Korea is heavily dependent on energy resources from other countries and nuclear energy accounts for 31% of Korea's electric power generation as a major energy. However, Korea has many limitations in uranium enrichment and spent fuel reprocessing under the current Korea-U.S. nuclear agreement, although they are economically and politically important to Korea due to a significant problems in nuclear fuel storages. Therefore, in this paper, we first examine those example countries – Japan, Vietnam, and Iran – that have made nuclear agreements with the U.S. or have changed their agreements to allow the enrichment of uranium and the reprocessing of spent fuel. Then, we analyze those countries' nuclear energy policies and review their strategic repositioning in the relationship with the U.S. We find that a strong political stance for peaceful usage of nuclear energy including the legislation of nuclear laws as was the case of Japan. In addition, it is important for Korea to acquire advanced technological capability such as sodium-cooled fast reactor (SFR) because SFR technologies require plutonium to be used as fuel rather than uranium-235. In addition, Korea needs to leverage its position in nuclear agreement between China and the U.S. as was the case of Vietnam

  12. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    International Nuclear Information System (INIS)

    Allen, K.J.; Bolshinsky, I.; Biro, L.L.; Budu, M.E.; Zamfir, N.V.; Dragusin, M.

    2010-01-01

    Romania safely air shipped 23.7 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel from the VVR-S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world's first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3. country under the RRRFR program and the 14. country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment. (authors)

  13. Cooking rice in excess water reduces both arsenic and enriched vitamins in the cooked grain.

    Science.gov (United States)

    Gray, Patrick J; Conklin, Sean D; Todorov, Todor I; Kasko, Sasha M

    2016-01-01

    This paper reports the effects of rinsing rice and cooking it in variable amounts of water on total arsenic, inorganic arsenic, iron, cadmium, manganese, folate, thiamin and niacin in the cooked grain. We prepared multiple rice varietals both rinsed and unrinsed and with varying amounts of cooking water. Rinsing rice before cooking has a minimal effect on the arsenic (As) content of the cooked grain, but washes enriched iron, folate, thiamin and niacin from polished and parboiled rice. Cooking rice in excess water efficiently reduces the amount of As in the cooked grain. Excess water cooking reduces average inorganic As by 40% from long grain polished, 60% from parboiled and 50% from brown rice. Iron, folate, niacin and thiamin are reduced by 50-70% for enriched polished and parboiled rice, but significantly less so for brown rice, which is not enriched.

  14. Latest developments in rolled fuels for materials-testing reactors: a trend towards the use of low-enriched uranium

    International Nuclear Information System (INIS)

    Fanjas, Y.

    1981-01-01

    The properties of rolled fuels and the work done in this field by CERCA is described. The technology developed conforms to low enrichment requirements, whilst guaranteeing a satisfactory level of reactor performance [fr

  15. 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.

  16. Russian-Origin Highly Enriched Uranium Spent Nuclear Fuel Shipment From Bulgaria

    International Nuclear Information System (INIS)

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

    2009-01-01

    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.

  17. Nuclear magnetic resonance spectroscopic investigation of anode exhaust of direct methanol fuel cells without isotope enrichment

    International Nuclear Information System (INIS)

    Byun, Young Seok; Hwang, Reo Yun; Han, Ochee

    2016-01-01

    Fuel cells are devices that electrochemically convert the chemical energy of fuels such as natural gas, gasoline, and methanol, into electricity. Fuel cells more efficiently use energy than internal combustion engines and do not produce undesirable pollutants, such as NO_x ,SO_x and particulates. Fuel cells can be distinguished from one another by their electrolytes. Among the various direct alcohol fuel cells, direct methanol fuel cells (DMFCs) have been developed most. However, DMFCs have several practical problems such as methanol crossove r from an anode to a cathode and slow methanol oxidation reaction rates. Therefore, understanding the electrochemical reaction mechanisms of DMFCs may provide clues to solve these problems, and various analytical methods have been employed to examine these mechanisms. We demonstrated that "1H and "1"3C NMR spectroscopy can be used for analyzing anode exhausts of DMFCs operated with methanol without any isotope enrichment. However, the low sensitivity of NMR spectroscopy hindered our efforts to detect minor reaction intermediates. Therefore, sensitivity enhancement techniques such as dynamic nuclear polarization (DNP) NMR methods and/or presaturation methods to increase the dynamic range of the proton spectra by pre-saturating large water signals, are expected to be useful to detect low-concentration species

  18. Nuclear magnetic resonance spectroscopic investigation of anode exhaust of direct methanol fuel cells without isotope enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Young Seok; Hwang, Reo Yun; Han, Ochee [Western Seoul Center, Korea Basic Science Institute, Seoul (Korea, Republic of)

    2016-12-15

    Fuel cells are devices that electrochemically convert the chemical energy of fuels such as natural gas, gasoline, and methanol, into electricity. Fuel cells more efficiently use energy than internal combustion engines and do not produce undesirable pollutants, such as NO{sub x} ,SO{sub x} and particulates. Fuel cells can be distinguished from one another by their electrolytes. Among the various direct alcohol fuel cells, direct methanol fuel cells (DMFCs) have been developed most. However, DMFCs have several practical problems such as methanol crossove r from an anode to a cathode and slow methanol oxidation reaction rates. Therefore, understanding the electrochemical reaction mechanisms of DMFCs may provide clues to solve these problems, and various analytical methods have been employed to examine these mechanisms. We demonstrated that {sup 1}H and {sup 13}C NMR spectroscopy can be used for analyzing anode exhausts of DMFCs operated with methanol without any isotope enrichment. However, the low sensitivity of NMR spectroscopy hindered our efforts to detect minor reaction intermediates. Therefore, sensitivity enhancement techniques such as dynamic nuclear polarization (DNP) NMR methods and/or presaturation methods to increase the dynamic range of the proton spectra by pre-saturating large water signals, are expected to be useful to detect low-concentration species.

  19. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

  20. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

    1979-04-01

    Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

  1. Timing of grazing to reduce cheatgrass fuels

    Science.gov (United States)

    The introduction and subsequent invasion of cheatgrass onto millions of acres of Great Basin rangelands has revolutionized secondary succession by providing a fine-textured early maturing fuel that has increased the chance, rate, spread and season of wildfires. With such vast acreages of landscapes ...

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

    International Nuclear Information System (INIS)

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    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 UO 2 and PuO 2 -UO 2 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 UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 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 235 U 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

  3. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  4. Volatile behaviour of enrichment uranium in the total nuclear fuel price

    International Nuclear Information System (INIS)

    Arnaiz, J.; Inchausti, J. M.; Tarin, F.

    2004-01-01

    In this article the historical high volatile behaviour of the total nuclear fuel price is evaluated quantitatively and it is concluded that it has been due mainly to the fluctuations of the price of the principal components of enriched uranium (concentrates and enrichment). In order to avoid the negative effects of this volatiles behaviour as far as possible, a basic strategy in the uranium procurement activities is recommended (union of buyers, diversification of supplier, stock management, optimisation of contract portfolio and suitable currency management that guarantees a reliable uranium supply at reasonable prices. These guidelines are those that ENUSA has been following on behalf of the Spanish Utilities in the Commission of Uranium Procurement (CAU in Spanish). (Author) 11 refs

  5. Environmental enrichment and exercise are better than social enrichment to reduce memory deficits in amyloid beta neurotoxicity.

    Science.gov (United States)

    Prado Lima, Mariza G; Schimidt, Helen L; Garcia, Alexandre; Daré, Letícia R; Carpes, Felipe P; Izquierdo, Ivan; Mello-Carpes, Pâmela B

    2018-03-06

    Recently, nongenetic animal models to study the onset and development of Alzheimer's disease (AD) have appeared, such as the intrahippocampal infusion of peptides present in Alzheimer amyloid plaques [i.e., amyloid-β (Aβ)]. Nonpharmacological approaches to AD treatment also have been advanced recently, which involve combinations of behavioral interventions whose specific effects are often difficult to determine. Here we isolate the neuroprotective effects of three of these interventions-environmental enrichment (EE), anaerobic physical exercise (AnPE), and social enrichment (SE)-on Aβ-induced oxidative stress and on impairments in learning and memory induced by Aβ. Wistar rats were submitted to 8 wk of EE, AnPE, or SE, followed by Aβ infusion in the dorsal hippocampus. Short-term memory (STM) and long-term memory (LTM) of object recognition (OR) and social recognition (SR) were evaluated. Biochemical assays determined hippocampal oxidative status: reactive oxygen species, lipid peroxidation by thiobarbituric acid reactive substance (TBARS) test, and total antioxidant capacity by ferric reducing/antioxidant power (FRAP), as well as acetylcholinesterase activity. Aβ infusion resulted in memory deficits and hippocampal oxidative damage. EE and AnPE prevented all memory deficits (STM and LTM of OR and SR) and lipid peroxidation (i.e., TBARS). SE prevented only the SR memory deficits and the decrease of total antioxidant capacity decrease (i.e., FRAP). Traditionally, findings obtained with EE protocols do not allow discrimination of the roles of the three individual factors involved. Here we demonstrate that EE and physical exercise have better neuroprotective effects than SE in memory deficits related to Aβ neurotoxicity in the AD model tested.

  6. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, Y.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.

    2012-01-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  7. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C. [Joint Inst. for Power and Nuclear Research-Sosny, 99 Academician A.K.Krasin Str, Minsk 220109 (Belarus)

    2012-07-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  8. An experimental prescribed burn to reduce fuel hazard in chaparral

    Science.gov (United States)

    Lisle R. Green

    1970-01-01

    The feasibility of reducing fuel hazard in chaparral during safe weather conditions was studied in an experimental prescribed burn in southern California. Burning was done under fuel and weather conditions when untreated brush would not bum readily. Preparatory treatment included smashing of brush on strips with a bulldozer, and reduction of moisture content of leaves...

  9. 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.

  10. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

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

    International Nuclear Information System (INIS)

    Reardon, P.T.; Mullen, M.F.; Harms, N.L.

    1981-02-01

    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

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

    International Nuclear Information System (INIS)

    Milne, J.M.; Lidington, B.; Hawker, B.M.

    1991-01-01

    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.)

  13. Proposal of new 235U nuclear data to improve keff biases on 235U enrichment and temperature for low enriched uranium fueled lattices moderated by light water

    International Nuclear Information System (INIS)

    Wu, Haicheng; Okumura, Keisuke; Shibata, Keiichi

    2005-06-01

    The under prediction of k eff depending on 235 U enrichment in low enriched uranium fueled systems, which had been a long-standing puzzle especially for slightly enriched ones, was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k eff underestimation vs. temperature increase, which was observed in the sightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of 235 U and 238 U, we propose a new evaluation of 235 U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of 235 U and the 238 U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems. (author)

  14. Reduced chemical kinetic mechanisms for hydrocarbon fuels

    International Nuclear Information System (INIS)

    Montgomery, C.J.; Cremer, M.A.; Heap, M.P.; Chen, J-Y.; Westbrook, C.K.; Maurice, L.Q.

    1999-01-01

    Using CARM (Computer Aided Reduction Method), a computer program that automates the mechanism reduction process, a variety of different reduced chemical kinetic mechanisms for ethylene and n-heptane have been generated. The reduced mechanisms have been compared to detailed chemistry calculations in simple homogeneous reactors and experiments. Reduced mechanisms for combustion of ethylene having as few as 10 species were found to give reasonable agreement with detailed chemistry over a range of stoichiometries and showed significant improvement over currently used global mechanisms. The performance of reduced mechanisms derived from a large detailed mechanism for n-heptane was compared to results from a reduced mechanism derived from a smaller semi-empirical mechanism. The semi-empirical mechanism was advantageous as a starting point for reduction for ignition delay, but not for PSR calculations. Reduced mechanisms with as few as 12 species gave excellent results for n-heptane/air PSR calculations but 16-25 or more species are needed to simulate n-heptane ignition delay

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

    International Nuclear Information System (INIS)

    Suripto, A.; Sugondo, S.; Nasution, H.

    1994-01-01

    The first low-enriched U 3 Si 2 -Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235 U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U 3 Si 2 -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

  16. Forest fuel reduces the nitrogen load

    International Nuclear Information System (INIS)

    Lundborg, A.

    1993-03-01

    A study of the literature was made on the basis of the following hypothesis: ''If nitrogen-rich felling residues are removed from the forest, the nitrogen load on the forest ecosystem is decreased and the risk of nitrogen saturation also decreases''. The study was designed to provide information on how the nitrogen situation is influenced if felling residues are removed from nitrogen-loaded forests and used as fuel. Felling residues release very little nitrogen during the first years after felling. They can immobilize nitrogen from the surroundings, make up a considerable addition to the nitrogen store in the soil, but also release nitrogen in later stages of degradation. The slash has an influence on the soil climate and thus on soil processes. Often there is an increase in the mineralization of litter and humus below the felling residues. At the same time, nitrification is favoured, particularly if the slash is left in heaps. Felling residues contain easily soluble nutrients that stimulate the metabolization of organic matter that otherwise is rather resistant to degradation. The slash also inhibits the clear-cut vegetation and its uptake of nitrogen. These effects result in increased leaching of nitrogen and minerals if the felling residues are left on the site. (99 refs.)

  17. Does a renewable fuel standard for biofuels reduce climate costs?

    Energy Technology Data Exchange (ETDEWEB)

    Greaker, Mads; Hoel, Michael; Rosendahl, Knut Einar

    2012-07-01

    Recent contributions have questioned whether biofuels policies actually lead to emissions reductions, and thus lower climate costs. In this paper we make two contributions to the literature. First, we study the market effects of a renewable fuel standard. Opposed to most previous studies we model the supply of fossil fuels taking into account that fossil fuels is a non-renewable resource. Second, we model emissions from land use change explicitly when we evaluate the climate effects of the renewable fuel standard. We find that extraction of fossil fuels most likely will decline initially as a consequence of the standard. Thus, if emissions from biofuels are sufficiently low, the standard will have beneficial climate effects. Furthermore, we find that the standard tends to reduce total fuel (i.e., oil plus biofuels) consumption initially. Hence, even if emissions from biofuels are substantial, climate costs may be reduced. Finally, if only a subset of countries introduce a renewable fuel standard, there will be carbon leakage to the rest of the world. However, climate costs may decline as global extraction of fossil fuels is postponed.(Author)

  18. Effects of fuel enrichment on the physics characteristics of plutonium-fueled light water high converter reactors

    International Nuclear Information System (INIS)

    Chawla, R.; Seiler, R.; Gmur, K.

    1986-01-01

    Investigations have been carried out for three additional cores of the phase 1 experimental program on light water high converter reactor test lattices in the PROTEUS facility. An 8% (average) fissile plutonium tight-pitch lattice with a fuel/moderator volumetric ratio of 2.0 was considered. As for the earlier reported 6% (average) fissile plutonium test lattice, H 2 O, Dowtherm, and air were the moderator state investigated. Significant enrichment-dependent trends have been identified in the comparisons of calculated and experimental results for the wet (moderated cases, particularly for the important reaction rate ratio of 238 U capture of 239 Pu fission. These are then reflected in the comparison of moderator voidage characteristics, expressed in terms of individual components of the kinfinity void coefficient

  19. Effects of fuel enrichment on the physics characteristics of plutonium-fueled light water high converter reactors

    International Nuclear Information System (INIS)

    Chawla, R.; Seiler, R.; Gmuer, K.

    1986-01-01

    Investigations have been carried out for three additional cores of the phase 1 experimental program on light water high converter reactor test lattices in the PROTEUS facility. An 8% (average) fissile plutonium tight-pitch lattice with a fuel/moderator volumetric ratio of 2.0 was considered. As for the earlier reported 6% (average) fissile plutonium test lattice, H 2 O, Dowtherm, and air were the moderator states investigated. Significant enrichment-dependent trends have been identified in the comparisons of calculated and experimental results for the wet (moderated) cases, particularly for the important reaction rate ratio of 238 U capture to 239 Pu fission. These are then reflected in the comparison of moderator voidage characteristics, expressed in terms of individual components of the k-infinity void coefficient. (author)

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

    International Nuclear Information System (INIS)

    Giorsetti, Domingo R.; Perez, Edmundo E.

    1983-01-01

    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: UAl x -Al and U 3 O 8 -Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm 3 for the UAI x -Al line and 2.4-3.0 g/cm 3 for the U 3 O 8 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/cm 3 in U 3 O 8 -Al fuel and of 2.4 g/cm 3 in UAI x -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/cm 3 with U 3 O 8 -Al and 2.52 g/cm 3 with UAl x -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 U 3 Si-Al. Three months later

  1. Conversion to low-enriched fuel in research reactor aspects of licensing the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Jacquemin, J.

    1985-01-01

    Conversion to low-enriched fuel and usage of new developed highly densified fuel in research-reactors will be an essential alteration in operating the reactor. According to the German Energy Act this has to be licensed. here might be some risk to the licensee of an older research-reactor by suspending his operating license because he cannot meet current requirements to be fulfilled or because of a court decision.Disposal of irradiated fuel elements of the new fuel type is a further significant problem which has to be solved before issuing a new license. (author)

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

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Kanda, Keiji; Shibata, Toshikazu

    1985-01-01

    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 x -Al, U 3 O 8 -Al and U 3 Si 2 -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. (author)

  3. Ramp metering with an objective to reduce fuel consumption

    OpenAIRE

    Vreeswijk, Jacob Dirk; Woldeab, Zeremariam; de Koning, Anne; Bie, Jing

    2011-01-01

    Ramp meters successfully decrease congestion but leave a burden on the traffic situation at on-ramps. Chaotic queuing leads to many stop-and-go movements and causes inefficiency where fuel consumption is concerned. As part of the eCoMove project, complementary strategies are being designed and evaluated to reduce fuel consumption at metered on-ramps, using vehicle-to-infrastructure communication. This paper presents the design of two strategies, as well as their effect as derived from simulat...

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

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G., E-mail: evanslg@ornl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T.; Menlove, Howard O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Schwalbach, Peter; Baere, Paul De [European Commission, Euratom Safeguards Office (Luxembourg); Browne, Michael C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-11-21

    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 (Gd{sub 2}O{sub 3}) 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 {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

  5. 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)

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

    2013-01-01

    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 (Gd 2 O 3 ) 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 241 AmLi (α,n) interrogation source strength of 5.7×10 4 s −1 . Furthermore, the calibration range of the new collar has been extended to verify 235 U content in variable PWR fuel designs in the presence of up to

  6. Neutronics and thermalhydraulics characteristics of the CANDU core fueled with slightly enriched uranium 0.9% U235

    International Nuclear Information System (INIS)

    Raica, V.; Sindile, A.

    1999-01-01

    The interest concerning the slightly enriched uranium (SEU) fuel cycle is due to the possibility to adapt (to convert) the current reactor design using natural uranium fuel to this cycle. Preliminary evaluations based on discharged fuel burnup estimates versus enrichment and on Canadian experience in fuel irradiation suggest that for a 0.93% U-235 enrichment no design modifications are required, not even for the fuel bundle. The purpose of this paper is to resume the results of the studies carried on in order to clarify this problem. The calculation methodology used in reactor physics and thermal-hydraulics analyses that were performed adapted and developed the AECL suggested methodology. In order to prove the possibility to use the SEU 0.93% without any design modification, all the main elements from the CANDU Reactor Physics Design Manual were studied. Also, some thermal-hydraulics analyses were performed to ensure that the operating and safety parameters were respected. The estimations sustain the assumption that the current reactor and fuel bundle design is compatible to the using of the SEU 0.93% fuel. (author)

  7. Study on usage of low enriched uranium Russian type fuel elements for design of an experimental ADS research reactor

    International Nuclear Information System (INIS)

    Pesic, M.P.

    2005-01-01

    Conceptual design of an accelerator driven sub-critical experimental research reactor (ADSRR) was initiated in 1999 at the Vinca Institute of Nuclear Sciences, Serbia and Montenegro. Initial results of neutronic analyses of the proposed ADSRR-H were carried out by Monte Carlo based codes and available high-enriched uranium dioxide (HEU) dispersed Russian type TVR-S fuel elements (FE) placed in a lead matrix. Beam of charged particles (proton or deuteron) would be extracted from the high-energy channel H5B of the VINCY cyclotron of the TESLA Accelerator Installation. In 2002, the Vinca Institute has, in compliance with the Reduced Enrichment for Research and Test Reactors (RERTR) Program, returned fresh HEU TVR-S type FEs back to the Russian Federation. Since usage of HEU FEs in research reactors is not further recommended, a new study of an ADSRR-L conceptual design has initiated in Vinca Institute in last two years, based on assumed availability of low-enriched uranium (LEU) dispersed type TVR-S FEs. Initial results of numerical simulations of this new ADSRR-L, published for the first time in this paper, shows that such a small low neutron flux system can be used as an experimental - 'demonstration' - ADS with neutron characteristics similar to proposed well-known lead moderated and cooled power sub-critical ADS with intermediate neutron spectrum. Neutron spectrum characteristics of the ADSRR-L are compared to ones of the ADSRR-H with the same mass (7.7 g) of 235 U nuclide per TVR-S FE. (author)

  8. 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.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    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.

  10. Return of 80% highly enriched uranium fresh fuel from Yugoslavia to Russia

    International Nuclear Information System (INIS)

    Pesic, M.; Sotic, O.; Subotic, K.; Hopwood, W. Jr; Moses, S.; Wander, T.; Smirnov, A.; Kanashov, B.; Eshcherkin, A.; Efarov, S.; Olivieri, C.; Loghin, N. E.

    2003-01-01

    The transport of almost 50 kg of highly enriched (80%) uranium (HEU), in the form of fresh TVR-S fuel elements, from the Vin a Institute of Nuclear Sciences, Yugoslavia, to the Russian Federation for uranium reprocessing was carried out in August 2002. This act was a contribution of the Government of the Federal Republics of Yugoslavia (now Serbia and Montenegro) to the world's joint efforts to prevent possible actions of terrorists against nuclear material that potentially would be usable for the production of nuclear weapons. Basic aspects of this complex operation, carried out mainly by transport teams of the Vinca Institute and of the Institute for Safe Transport of Nuclear Materials from Dimitrovgrad, Russian Federation, are described in this paper. A team of IAEA safety inspectors and experts from the DOE, USA, for transport and non-proliferation, supported the whole operation. (author)

  11. 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.

  12. Return of 80% highly enriched uranium fresh fuel from Yugoslavia to Russia

    International Nuclear Information System (INIS)

    Pesic, M.; Sotic, O.; Subotic, K.; Hopwood, W. Jr; Moses, S.; Wander, T.; Smirnov, A.; Kanashov, B.; Eshcherkin, A.; Efarov, S.; Olivieri, C.; Loghin, N. E.

    2003-01-01

    The transport of almost 50 kg of highly enriched (80%) uranium (HEU), in the form of fresh TVR-S fuel elements, from the Vinca Institute of Nuclear Sciences, Yugoslavia, to the Russian Federation for uranium reprocessing was carried out in August 2002. This act was a contribution of the Government of the Federal Republics of Yugoslavia (now Serbia and Montenegro) to the world's joint efforts to prevent possible actions of terrorists against nuclear material that potentially would be usable for the production of nuclear weapons. Basic aspects of this complex operation, carried out mainly by transport teams of the Vinca Institute and of the Institute for Safe Transport of Nuclear Materials from Dimitrovgrad, Russian Federation, are described in this paper. A team of IAEA safety inspectors and experts from the DOE, USA, for transport and non-proliferation, supported the whole operation. (author)

  13. Determination of burnup, cooling time and initial enrichment of PWR spent fuel by use of gamma-ray activity ratios

    International Nuclear Information System (INIS)

    Min, D.K.; Park, H.J.; Park, K.J.; Ro, S.G.; Park, H.S.

    1999-01-01

    The Korea Atomic Energy Institute has been developing the algorithms for sequential determination of cooling time, initial enrichment and burnup of the PWR spent fuel assembly by use of gamma ratio measurements, i.e. 134 Cs/ 137 Cs, 154 Eu/ 137 Cs and 106 Ru 137 Cs/( 134 Cs) 2 . Calculations were performed by applying the ORIGEN-S code. This method has advantages over combination techniques of neutron and gamma measurement, because of its simplicity and insensitivity to the measurement geometry. For verifying the algorithms an experiment for determining the cooling time, initial enrichment and burnup of the two PWR spent fuel rods was conducted by use of high-resolution gamma detector (HPGe) system only. This paper describes the method used and interim results of the experiment. This method can be applied for spent fuel characterization, burnup credit and safeguards of the spent fuel management facility

  14. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    International Nuclear Information System (INIS)

    Uriarte, A.; Ramos, L.; Estrada, J.; del Val, J. L.

    1962-01-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO 2 F 2 solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs

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

    International Nuclear Information System (INIS)

    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 ( 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 20 fissions, the number of fissions equals about 10 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

  16. 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.

  17. Progress of the RERTR [Reduced Enrichment Research and Test Reactor] Program in 1989

    International Nuclear Information System (INIS)

    Travelli, A.

    1989-01-01

    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 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs

  18. Enrichment: CRISLA [chemical reaction by isotope selective activation] aims to reduce costs

    International Nuclear Information System (INIS)

    Eerkens, J.W.

    1989-01-01

    Every year, more than $3 billion is spent on enriching uranium. CRISLA (Chemical Reaction by Isotope Selective Activation) uses a laser-catalyzed chemical reaction which, its proponents claim, could substantially reduce these costs. In CRISLA, an infrared CO laser illuminates the intracavity reaction cell (IC) at a frequency tuned to excite primarily UF 6 . When UF 6 and co-reactant RX are passed through the IC, the tuned laser photons preferentially enhance the reaction of UF 6 with RX ten-thousand-fold over the thermal reaction rate. Thus the laser serves as an activator and the chemical energy for separation is largely chemical. (author)

  19. 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.

  20. Bioethanol fuel production from rambutan fruit biomass as reducing ...

    African Journals Online (AJOL)

    The depletion of fossil fuels impacts on the increase of petroleum price and has triggered the finding of alternative and renewable energy. Biofuel has attracted the attention of researchers all over the world due to reducing the environmental impacts of elevated carbon monoxide. Abundant of fruits waste can be reused in the ...

  1. Safety concerning the alteration in fuel material usage (new installation of the uranium enrichment pilot plant) at Ningyo Pass Mine of Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    A report of the Committee on Examination of Nuclear Fuel Safety was presented to the Atomic Energy Commission of Japan, which is concerned with the safety in the alteration of fuel material usage (new installation of the uranium enrichment pilot plant) at the Ningyo Pass Mine. Its safety was confirmed. The alteration, i.e. installation of the uranium enrichment pilot plant, is as follows. Intended for the overall test of centrifugal uranium enrichment technology, the pilot plant includes a two-storied main building of about 9,000 m 2 floor space, containing centrifuges, UF 6 equipment, etc., a uranium storage of about 1,000 m 2 floor space, and a waste water treatment facility, two-storied with about 300 m 2 floor space. The contents of the examination are safety of the facilities, criticality control, radiation control, waste treatment, and effects of accidents on the surrounding environment. (Mori, K

  2. Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

    Energy Technology Data Exchange (ETDEWEB)

    Chambers, Alex, E-mail: acchamb@gmail.com; Ragusa, Jean C., E-mail: jean.ragusa@tamu.edu

    2014-04-01

    Highlights: • Study of multiple recycling passes of transuranic elements: (a) without exceeding 5 wt.% on U-235 enrichment; (b) using PWR fuel assemblies compatible with current reactor core internals. • Isotopic concentrations tend towards an equilibrium after 15 recycle passes, suggesting that thermal recycling may be continued beyond that point. • Radiotoxicity comparisons for once-through UOX, once-recycle MOX-Pu, and multiple recycle passes of MOX-PuNpAm and MOX-PuNpAmCm are presented. - Abstract: This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: {sup 235}U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes.

  3. Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

    International Nuclear Information System (INIS)

    Chambers, Alex; Ragusa, Jean C.

    2014-01-01

    Highlights: • Study of multiple recycling passes of transuranic elements: (a) without exceeding 5 wt.% on U-235 enrichment; (b) using PWR fuel assemblies compatible with current reactor core internals. • Isotopic concentrations tend towards an equilibrium after 15 recycle passes, suggesting that thermal recycling may be continued beyond that point. • Radiotoxicity comparisons for once-through UOX, once-recycle MOX-Pu, and multiple recycle passes of MOX-PuNpAm and MOX-PuNpAmCm are presented. - Abstract: This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: 235 U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes

  4. Reducing Fuel Volatility. An Additional Benefit From Blending Bio-fuels?

    Energy Technology Data Exchange (ETDEWEB)

    Bailis, R. [Yale School of Forestry and Environmental Studies, 195 Prospect Street, New Haven, CT 06511 (United States); Koebl, B.S. [Utrecht University, Science Technology and Society, Budapestlaan 6, 3584 CD Utrecht (Netherlands); Sanders, M. [Utrecht University, Utrecht School of Economics, Janskerkhof 12, 3512 BL Utrecht (Netherlands)

    2011-02-15

    Oil price volatility harms economic growth. Diversifying into different fuel types can mitigate this effect by reducing volatility in fuel prices. Producing bio-fuels may thus have additional benefits in terms of avoided damage to macro-economic growth. In this study we investigate trends and patterns in the determinants of a volatility gain in order to provide an estimate of the tendency and the size of the volatility gain in the future. The accumulated avoided loss from blending gasoline with 20 percent ethanol-fuel estimated for the US economy amounts to 795 bn. USD between 2010 and 2019 with growing tendency. An amount that should be considered in cost-benefit analysis of bio-fuels.

  5. Environmental enrichment reduces chronic psychosocial stress-induced anxiety and ethanol-related behaviors in mice.

    Science.gov (United States)

    Bahi, Amine

    2017-07-03

    Previous research from our laboratory has shown that exposure to chronic psychosocial stress increased voluntary ethanol consumption and preference as well as acquisition of ethanol-induced conditioned place preference (CPP) in mice. This study was done to determine whether an enriched environment could have "curative" effects on chronic psychosocial stress-induced ethanol intake and CPP. For this purpose, experimental mice "intruders" were exposed to the chronic subordinate colony (CSC) housing for 19 consecutive days in the presence of an aggressive "resident" mouse. At the end of that period, mice were tested for their anxiety-like behavior using the elevated plus maze (EPM) test then housed in a standard or enriched environment (SE or EE respectively). Anxiety and ethanol-related behaviors were investigated using the open field (OF) test, a standard two-bottle choice drinking paradigm, and the CPP procedure. As expected, CSC exposure increased anxiety-like behavior and reduced weight gain as compared to single housed colony (SHC) controls. In addition, CSC exposure increased voluntary ethanol intake and ethanol-CPP. Interestingly, we found that EE significantly and consistently reduced anxiety and ethanol consumption and preference. However, neither tastants' (saccharin and quinine) intake nor blood ethanol metabolism were affected by EE. Finally, and most importantly, EE reduced the acquisition of CPP induced by 1.5g/kg ethanol. Taken together, these results support the hypothesis that EE can reduce voluntary ethanol intake and ethanol-induced conditioned reward and seems to be one of the strategies to reduce the behavioral deficits and the risk of anxiety-induced alcohol abuse. Copyright © 2017 Elsevier Inc. All rights reserved.

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

    International Nuclear Information System (INIS)

    SCHWINKENDORF, K.N.

    2006-01-01

    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 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 supply useful

  7. Instruments for reducing the specific fuel consumption of cars

    International Nuclear Information System (INIS)

    Hammer, S.; Maibach, M.; Marti, P.

    2001-01-01

    This report for the Swiss Federal Office of Energy (SFOE) presents three possible courses of action that are to be taken to reduce the specific fuel consumption of private cars. The report first examines existing targets and the degree to which they have been met up to now, whereby the situation both in Switzerland and in the European Union is looked at. The report makes a suggestion for a future target scenario and elaborates three possible ways to met these targets: regulations on fuel-consumption, a bonus/malus system and tradable certificates. For each of the proposed instruments, the report examines implementation variants and discusses the means for their implementation. The report presents the best models for each of the implementation-variants on the basis of comparisons and the results of evaluations of their effects. For these chosen variants, the authors present comparisons of their effect on fuel consumption in graphical form and recommend tradable certificates as the best instrument

  8. Effect of Hydrogen and Hydrogen Enriched Compressed Natural Gas Induction on the Performance of Rubber Seed Oil Methy Ester Fuelled Common Rail Direct Injection (CRDi Dual Fuel Engines

    Directory of Open Access Journals (Sweden)

    Mallikarjun Bhovi

    2017-06-01

    Full Text Available Renewable fuels are in biodegradable nature and they tender good energy security and foreign exchange savings. In addition they address environmental concerns and socio-economic issues. The present work presents the experimental investigations carried out on the utilization of such renewable fuel combinations for diesel engine applications. For this a single-cylinder four-stroke water cooled direct injection (DI compression ignition (CI engine provided with CMFIS (Conventional Mechanical Fuel Injection System was rightfully converted to operate with CRDi injection systems enabling high pressure injection of Rubber seed oil methyl ester (RuOME in the dual fuel mode with induction of varied gas flow rates of hydrogen and hydrogen enriched CNG (HCNG gas combinations. Experimental investigations showed a considerable improvement in dual fuel engine performance with acceptable brake thermal efficiency and reduced emissions of smoke, hydrocarbon (HC, carbon monoxide (CO and slightly increased nitric oxide (NOx emission levels for increased hydrogen and HCNG flow rates. Further CRDi facilitated dual fuel engine showed improved engine performance compared to CMFIS as the former enabled high pressure (900 bar injection of the RuOME and closer to TDC (Top Dead Centre as well. Combustion parameters such as ignition delay, combustion duration, pressure-crank angle and heat release rates were analyzed and compared with baseline data generated. Combustion analysis showed that the rapid rate of burning of hydrogen and HCNG along with air mixtures increased due to presence of hydrogen in total and in partial combination with CNG which further resulted into higher cylinder pressures and energy release rates. However, sustained research that can provide feasible engine technology operating on such fuels in dual fuel operation can pave the way for continued fossil fuel usage.

  9. Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

    1996-12-01

    The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

  10. Feasibility Study on Nitrogen-15 Enrichment and Recycling System for Innovative FR Cycle System With Nitride Fuel

    International Nuclear Information System (INIS)

    Masaki Inoue; Kiyoshi Ono; Tsuna-aki Fujioka; Koji Sato; Takeo Asaga

    2002-01-01

    Highly-isotopically-enriched nitrogen (HE-N 2 ; 15 N abundance 99.9%) is indispensable for a nitride fueled fast reactor (FR) cycle to minimize the effect of carbon-14 ( 14 C) generated mainly by 14 N(n,p) 14 C reaction in the core on environmental burden. Thus, the development of inexpensive 15 N enrichment and recycling technology is one of the key aspects for the commercialization of a nitride fueled FR cycle. Nitrogen isotope separation by the gas adsorption technique was experimentally confirmed in order to obtain its technological perspective. A conventional pressure swing adsorption technique, which is already commercialized for recovering the nitrogen gas from multi-composition gas-mixture, would be suitable for recovering in both reprocessing and fuel fabrication to recycle the HE-N 2 gas. A couple of the nitride fuel cycle system concepts including the reprocessing and fuel fabrication process flow diagrams with the HE-N 2 gas recycling were newly designed for both aqueous and non-aqueous (pyrochemical) nitride fuel recycle plants, and also the effect of the HE-N 2 gas recycling on the economics of each concept was evaluated. (authors)

  11. Conversion of the RB reactor neutrons by highly enriched uranium fuel and lithium deuteride

    International Nuclear Information System (INIS)

    Strugar, P.; Sotic, O.; Ninkovic, M.; Pesic, M.; Altiparmakov, D.

    1981-01-01

    A thermal-to-fast-neutron converter has been constructed at the RB reactor. The material used for the conversion of thermal neutrons is highly enriched uranium fuel of Soviet production applied in Yugoslav heavy water experimental reactors RA and RB. Calculations and preliminary measurements show that the spectrum of converted neutrons only slightly differs from that of fission neutrons. The basic characteristics of converted neutrons can be expressed by the neutron radiation dose of 800 rad (8 Gy) for 1 h of reactor operation at a power level of 1 kW. This dose is approximately 10 times higher than the neutron dose at the same place without converter. At the same time, thermal neutron and gamma radiation doses are negligible. The constructed neutron converter offers wide possibilities for applications in reactor and nuclear physics and similar disciplines, where neutron spectra of high energies are required, as well as in the domain of neutron dosimetry and biological irradiations in homogeneous fields of larger dimensions. The possibility of converting thermal reactor neutrons with energies of about 14 MeV with the aid of lithium deuteride from natural lithium has been considered too. (author)

  12. recovery of enriched uranium from waste solution obtained from fuel fabrication laboratories

    International Nuclear Information System (INIS)

    Othman, S.H.A.

    2003-01-01

    reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of uranium (19.7% enriched with 235 U) from highly impure solution . the processing of uranium compounds for atomic energy project especially in FMPP(Egyptian 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-ethilene (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%. A mathematical model was suggested to stimulate the sorption process of uranium ions (or any other ion ) by column of solvent impregnated resin containing organic extractant (the same as the previous column) . An excellent agreement was founded between the experimental results and the mathematical model

  13. Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Carter, R.E.

    1985-01-01

    During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

  14. Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carter, R E

    1985-07-01

    During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

  15. Effect of Long Time Oxygen Exposure on Power Generation of Microbial Fuel Cell with Enriched Mixed Culture

    International Nuclear Information System (INIS)

    Mimi Hani Abu Bakar; Mimi Hani Abu Bakar; Mimi Hani Abu Bakar; Pasco, N.F.; Gooneratne, R.; Hong, K.B.; Hong, K.B.; Hong, K.B.

    2016-01-01

    In this study, we are interested in the effect of long time exposure of the microbial fuel cells (MFCs) to air on the electrochemical performance. Here, MFCs enriched using an effluent from a MFC operated for about eight months. After 30 days, the condition of these systems was reversed from aerobic to anaerobic and vice versa, and their effects were observed for 11 days. The results show that for anaerobic MFCs, power generation was reduced when the anodes were exposed to dissolved oxygen of 7.5 ppm. The long exposure of anodic biofilm to air led to poor electrochemical performance. The power generation recovered fully when air supply stopped entering the anode compartment with a reduction of internal resistance up to 53 %. The study was able to show that mixed facultative microorganism able to strive through the aerobic condition for about a month at 7.5 ppm oxygen or less. The anaerobic condition was able to turn these microbes into exoelectrogen, producing considerable power in relative to their aerobic state. (author)

  16. US enrichment reduction studies

    International Nuclear Information System (INIS)

    1979-06-01

    A major national program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is currently under way in the U.S., centered at the Argonne National Laboratory (ANL), to reduce the potential of research and test reactor fuels for increasing the proliferation of nuclear explosive devices. The main objective of the program is to provide the technical means by which the uranium enrichment to be used in these reactors can be reduced to less than 20% without significant economic and performance penalties. The criteria, basis and goals of the program are consistent with the results of a number of case studies which have been performed as part of the program

  17. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  18. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Uriarte, A; Ramos, L; Estrada, J; Val, J L. del

    1962-07-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO{sub 2}F{sub 2} solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs.

  19. The use of fuel of various enrichment for flux shaping; Koriscenje goriva razlicitog obogacenja za dobijanje zeljene raspodele neutronskog fluksa

    Energy Technology Data Exchange (ETDEWEB)

    Zavaljevski, N; Pesic, M; Strugar, P [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1980-07-01

    Spatial flux shaping, particularly obtaining maximum thermal neutron flux in experimental channels of a research reactor or flux flattening in a power reactor, is often desired in nuclear reactor utilization. Some experimental results of flux shaping at the RB reactor by use of the fuel of various enrichment are resented. Considerable increases in thermal neutron flux in central experimental channels is obtained and can serve as a starting point for further investigations as well as for comparison with theoretical models. (author)

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

    International Nuclear Information System (INIS)

    Ali Khan, L.; Ahmad, N.

    2002-01-01

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

  1. 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

  2. The main conditions ensured problemless implementation of 235U high enriched fuel in Kozloduy NPP (Bulgaria) - WWER-1000 Units

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.; Minkova, K.; Michaylov, G.; Penev, P.; Gerchev, N.

    2009-01-01

    The collected water chemistry and radiochemistry data during the operation of the Kozloduy NPP Unit 5 for the period 2006-2009 (12-th, 13-th 14-th and 15-th fuel cycles) undoubtedly indicate for WWER-1000 Units (whose specific features are: Steam generators with austenitic stainless steel 08Cr18N10T tubing; Steam generators are with horizontal straight tubing and Fuel elements cladding material is Zr-1%Nb (Zr1Nb) alloy), that one realistic way for problemless implementation of 235 U high enriched fuel have been found. The main feature characteristics of this way are: Implementation of solid neutron burnable absorbers together with the dissolved in coolant neutron absorber - natural boric acid; Application of fuel cladding materials with enough corrosion resistance by the specific fuel cladding environment created by presence of SNB; Keeping of suitable coolant water chemistry which ensures low corrosion rates of core- and out-of-core- materials and limits in core (cladding) depositions and restricts out-of-core radioactivity buildup. The realization of this way in WWER-1000 Units in Kozloduy NPP was practically carried out through: 1) Implementation of Russian fuel assemblies TVSA which have as fuel cladding material E-110 alloy (Zr1Nb) with enough high corrosion resistance by presence of sub-cooled nucleate boiling (SNB) and use burnable absorber (Gd) integrated in the uranium-gadolinium (U-Gd 2 O 3 ) fuel (fuel rod with 5.0% Gd 2 O 3 ); 2) Development and implementation of water chemistry primary circuit guidelines, which require the relation between boric acid concentration and total alkalising agent concentrations to ensure coolant pH 300 = 7.0 - 7.2 values during the whole operation period. The above mentioned conditions by the passing of WWER-1000 Units in NPP Kozloduy to uranium fuel with 4.4% 235 U (TVSA fuel assemblies) practically ensured avoidance of the creation of the necessary conditions for AOA onset. The operational experience (2006-2009) of the

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

    International Nuclear Information System (INIS)

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

    2007-01-01

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are

  5. Aircraft Engine Technology for Green Aviation to Reduce Fuel Burn

    Science.gov (United States)

    Hughes, Christopher E.; VanZante, Dale E.; Heidmann, James D.

    2013-01-01

    The NASA Fundamental Aeronautics Program Subsonic Fixed Wing Project and Integrated Systems Research Program Environmentally Responsible Aviation Project in the Aeronautics Research Mission Directorate are conducting research on advanced aircraft technology to address the environmental goals of reducing fuel burn, noise and NOx emissions for aircraft in 2020 and beyond. Both Projects, in collaborative partnerships with U.S. Industry, Academia, and other Government Agencies, have made significant progress toward reaching the N+2 (2020) and N+3 (beyond 2025) installed fuel burn goals by fundamental aircraft engine technology development, subscale component experimental investigations, full scale integrated systems validation testing, and development validation of state of the art computation design and analysis codes. Specific areas of propulsion technology research are discussed and progress to date.

  6. Fossil fuel derivatives with reduced carbon. Phase I final report

    Energy Technology Data Exchange (ETDEWEB)

    Kennel, E.B.; Zondlo, J.W.; Cessna, T.J.

    1999-06-30

    This project involves the simultaneous production of clean fossil fuel derivatives with reduced carbon and sulfur, along with value-added carbon nanofibers. This can be accomplished because the nanofiber production process removes carbon via a catalyzed pyrolysis reaction, which also has the effect of removing 99.9% of the sulfur, which is trapped in the nanofibers. The reaction is mildly endothermic, meaning that net energy production with real reductions in greenhouse emissions are possible. In Phase I research, the feasibility of generating clean fossil fuel derivatives with reduced carbon was demonstrated by the successful design, construction and operation of a facility capable of utilizing coal as well as natural gas as an inlet feedstock. In the case of coal, for example, reductions in CO{sub 2} emissions can be as much as 70% (normalized according to kilowatts produced), with the majority of carbon safely sequestered in the form of carbon nanofibers or coke. Both of these products are value-added commodities, indicating that low-emission coal fuel can be done at a profit rather than a loss as is the case with most clean-up schemes. The main results of this project were as follows: (1) It was shown that the nanofiber production process produces hydrogen as a byproduct. (2) The hydrogen, or hydrogen-rich hydrocarbon mixture can be consumed with net release of enthalpy. (3) The greenhouse gas emissions from both coal and natural gas are significantly reduced. Because coal consumption also creates coke, the carbon emission can be reduced by 75% per kilowatt-hour of power produced.

  7. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    International Nuclear Information System (INIS)

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

  8. Reducing stress and fuel consumption providing road information

    Directory of Open Access Journals (Sweden)

    Víctor CORCOBA MAGAÑA

    2014-12-01

    Full Text Available In this paper, we propose a solution to reduce the stress level of the driver, minimize fuel consumption and improve safety. The system analyzes the driving style and the driver’s workload during the trip while driving. If it discovers an area where the stress increases and the driving style is not appropriate from the point of view of energy efficiency and safety for a particular driver, the location of this area is saved in a shared database. On the other hand, the implemented solution warns a particular user when approaching a region where the driving is difficult (high fuel consumption and stress using the shared database based on previous recorded knowledge of similar drivers in that area. In this case, the proposal provides an optimal deceleration profile if the vehicle speed is not adequate. Therefore, he or she may adjust the vehicle speed with both a positive impact on the driver workload and fuel consumption. The Data Envelopment Analysis algorithm is used to estimate the efficiency of driving and the driver’s workload in in each area. We employ this method because there is no preconceived form on the data in order to calculate the efficiency and stress level. A validation experiment has been conducted using both a driving simulator and a real environment with 12 participants who made 168 driving tests. The system reduced the slowdowns (38%, heart rate (4.70%, and fuel consumption (12.41% in the real environment. The proposed solution is implemented on Android mobile devices and does not require the installation of infrastructure on the road. It can be installed on any model of vehicle.

  9. Nitrogen enriched combustion of a natural gas internal combustion engine to reduce NO.sub.x emissions

    Science.gov (United States)

    Biruduganti, Munidhar S.; Gupta, Sreenath Borra; Sekar, R. Raj; McConnell, Steven S.

    2008-11-25

    A method and system for reducing nitrous oxide emissions from an internal combustion engine. An input gas stream of natural gas includes a nitrogen gas enrichment which reduces nitrous oxide emissions. In addition ignition timing for gas combustion is advanced to improve FCE while maintaining lower nitrous oxide emissions.

  10. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    International Nuclear Information System (INIS)

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses

  11. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses.

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

    International Nuclear Information System (INIS)

    Schneider, R.A.; Nilson, R.; Jaech, J.L.

    1978-01-01

    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

  13. Blueberry polyphenol-enriched soybean flour reduces hyperglycemia, body weight gain and serum cholesterol in mice

    Science.gov (United States)

    Roopchand, Diana E.; Kuhn, Peter; Rojo, Leonel E.; Lila, Mary Ann; Raskin, Ilya

    2013-01-01

    Defatted soybean flour (DSF) can sorb and concentrate blueberry anthocyanins and other polyphenols, but not sugars. In this study blueberry polyphenol-enriched DSF (BB-DSF) or DSF were incorporated into very high fat diet (VHFD) formulations and provided ad libitum to obese and hyperglycemic C57BL/6 mice for 13 weeks to investigate anti-diabetic effects. Compared to the VHFD containing DSF, the diet supplemented with BB-DSF reduced weight gain by 5.6%, improved glucose tolerance, and lowered fasting blood glucose levels in mice within 7 weeks of intervention. Serum cholesterol of mice consuming the BB-DSF-supplemented diet was 13.2% lower than mice on the diet containing DSF. Compounds were eluted from DSF and BB-DSF for in vitro assays of glucose production and uptake. Compared to untreated control, doses of BB-DSF eluate containing 0.05 – 10 μg/μL of blueberry anthocyanins significantly reduced glucose production by 24% - 74% in H4IIE rat hepatocytes, but did not increase glucose uptake in L6 myotubes. The results indicate that delivery of blueberry polyphenols stabilized in a high-protein food matrix may be useful for the dietary management of pre-diabetes and/or diabetes. PMID:23220243

  14. Development of WWER-440 fuel. Use of fuel assemblies of 2-nd and 3-rd generations with increased enrichment

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Lushin, V.; Ananev, U.; Baranov, A.; Kukushkin, U.

    2009-01-01

    The problem of increasing the power of units at NPPs with WWER-440 is of current importance. There are all the necessary prerequisites for the above-stated problem as a result of updating the design of fuel assemblies and codes. The decrease of power peaking factor in the core is achieved by using profiled fuel assemblies, fuel-integrated burning absorber, FAs with modernized docking unit, modern codes, which allows decreasing conservatism of RP safety substantiation. A wide range of experimental studies of fuel behaviour has been performed which has reached burn-up of (50-60) MW·day/kgU in transition and emergency conditions, post-reactor studies of fuel assemblies, fuel rods and fuel pellets with a 5-year operating period have been performed, which prove high reliability of fuel, presence of a large margin in the fuel pillar, which helps reactor operation at increased power. The results of the work performed on introduction of 5-6 fuel cycles show that the ultimate fuel state on operability in WWER-440 reactors is far from being achieved. Neutron-physical and thermal-hydraulic characteristics of the cores of working power units with RP V-213 are such that actual (design and measured) power peaking factors on fuel assemblies and fuel rods, as a rule, are smaller than the maximum design values. This factor is a real reserve for power forcing. There is experience of operating Units 1, 2, 4 of the Kola NPP and Unit 2 of the Rovno NPP at increased power. Units of the Loviisa NPP are operated at 109 % power. During transfer to work at increased power it is reasonable to use fuel assemblies with increased height of the fuel pillar, which allows decreasing medium linear power distribution. Further development of the 2-nd generation fuel assembly design and consequent transition to working fuel assemblies of the 3-rd generation provides significant improvement of fuel consumption under the conditions of WWER-440 reactors operation with more continuous fuel cycles and

  15. Phylogenetic and functional diversity within toluene-degrading, sulphate-reducing consortia enriched from a contaminated aquifer.

    Science.gov (United States)

    Kuppardt, Anke; Kleinsteuber, Sabine; Vogt, Carsten; Lüders, Tillmann; Harms, Hauke; Chatzinotas, Antonis

    2014-08-01

    Three toluene-degrading microbial consortia were enriched under sulphate-reducing conditions from different zones of a benzene, toluene, ethylbenzene and xylenes (BTEX) plume of two connected contaminated aquifers. Two cultures were obtained from a weakly contaminated zone of the lower aquifer, while one culture originated from the highly contaminated upper aquifer. We hypothesised that the different habitat characteristics are reflected by distinct degrader populations. Degradation of toluene with concomitant production of sulphide was demonstrated in laboratory microcosms and the enrichment cultures were phylogenetically characterised. The benzylsuccinate synthase alpha-subunit (bssA) marker gene, encoding the enzyme initiating anaerobic toluene degradation, was targeted to characterise the catabolic diversity within the enrichment cultures. It was shown that the hydrogeochemical parameters in the different zones of the plume determined the microbial composition of the enrichment cultures. Both enrichment cultures from the weakly contaminated zone were of a very similar composition, dominated by Deltaproteobacteria with the Desulfobulbaceae (a Desulfopila-related phylotype) as key players. Two different bssA sequence types were found, which were both affiliated to genes from sulphate-reducing Deltaproteobacteria. In contrast, the enrichment culture from the highly contaminated zone was dominated by Clostridia with a Desulfosporosinus-related phylotype as presumed key player. A distinct bssA sequence type with high similarity to other recently detected sequences from clostridial toluene degraders was dominant in this culture. This work contributes to our understanding of the niche partitioning between degrader populations in distinct compartments of BTEX-contaminated aquifers.

  16. Remarks on the influence of enrichment reduction on fuel cycle costs

    International Nuclear Information System (INIS)

    Krull, W.

    1985-01-01

    The cost factors influencing the fuel cycle cost analysis for research reactors are discussed in detail with special emphasis on fuel element fabrication costs, burnup and reprocessing costs. Two different aspects for the conversion from HEU to LEU are considered: plus 14% U-235 weight per LEU fuel element and plus ca. 50 % U-235 weight per LEU fuel element. The cost factors and these conversion aspects were taken for calculating the changes in fuel cycle costs for the three different meat materials U 3 O 8 , U 3 Si 2 and U 3 Si. The results of these calculations can be summarized as following: - if in the HEU case the fuel loading and the burnup of a fuel element is low there will be some economic advantages in the LEU case; - if in the HEU case the fuel loading and the burnup of a fuel element is high there will be economic disadvantages in the LEU case. (author)

  17. Cooperative Russian-French experiment on plutonium-enriched fuels for fast burner reactor

    International Nuclear Information System (INIS)

    Zabud'ko, L.M.; Kurina, I.A.; Men'shikova, T.S.; Rogozkin, B.D.; Maershin, A.A.; Langi, A.; Pillon, S.

    2001-01-01

    Various kinds of nuclear fuels with an increased plutonium content are under study according to the program including three stages: fabrication, irradiation in BOR-60 reactor, post-irradiation examination. Flowsheets for fabricating pelletized and vibrocompacted fuels of UPu 0.45 O 2 , UPu 0.45 N, UPu 0.6 N, PuN + ZrN, PuO 2 + MgO are presented along with basic fuel properties. The irradiation of oxide fuel is carried out in an individual irradiation device at rated maximum temperature of the fuel at the beginning of irradiation equal to 2100 deg C. The irradiation of nitride fuel and the fuel based on inert matrices is performed in the other device with the aim of limitation of maximum temperature by the value of 1550 deg C. The duration of irradiation for all fuel types constitutes 750 EFPD. Fuel element charge in Bor-60 reactor core was realized in 2000 [ru

  18. Front end of the nuclear fuel cycle: options to reduce the risks of terrorism and proliferation

    International Nuclear Information System (INIS)

    Greenberg, E.V.C.; Hoenig, M.M.

    1987-01-01

    The authors' assessment of the prospects for advanced front end technologies and fuel assurances becoming effective mechanisms for achieving nonproliferation and antiterrorism objectives is relatively pessimistic unless they are integrated with back end accommodations such as the return of spent fuel. They recommend that further examination of front end assurances be linked to that accommodation. To be sure, certain real technological improvements may postpone the day when commercial use of nuclear explosive fuels, with all their attendant terrorism and proliferation risks, is justified. Indeed, improvements in LWRs, using well-understood technology combined with advanced enrichment techniques, could reduce uranium requirements up to 45% at the beginning of the next century and up to 30% a decade earlier, provided the economic and security incentives are present. On the institutional side, existing supply conditions put little pressure on importing countries to seek long-term supply assurances. Moreover, the political obstacles to creating new international institutions or arrangements are exceedingly difficult to overcome, especially without a heightened consciousness of the growing risks of civilian explosive nuclear materials and the political will to make these risks a high priority. 2 tables

  19. Reduced Toxicity Fuel Satellite Propulsion System Including Fuel Cell Reformer with Alcohols Such as Methanol

    Science.gov (United States)

    Schneider, Steven J. (Inventor)

    2001-01-01

    A reduced toxicity fuel satellite propulsion system including a reduced toxicity propellant supply for consumption in an axial class thruster and an ACS class thruster. The system includes suitable valves and conduits for supplying the reduced toxicity propellant to the ACS decomposing element of an ACS thruster. The ACS decomposing element is operative to decompose the reduced toxicity propellant into hot propulsive gases. In addition the system includes suitable valves and conduits for supplying the reduced toxicity propellant to an axial decomposing element of the axial thruster. The axial decomposing element is operative to decompose the reduced toxicity propellant into hot gases. The system further includes suitable valves and conduits for supplying a second propellant to a combustion chamber of the axial thruster, whereby the hot gases and the second propellant auto-ignite and begin the combustion process for producing thrust.

  20. Production and characterisation of reduced-fat and PUFA-enriched Burrata cheese.

    Science.gov (United States)

    Trani, Antonio; Gambacorta, Giuseppe; Gomes, Tommaso F; Loizzo, Pasqua; Cassone, Angela; Faccia, Michele

    2016-05-01

    Burrata is an Italian fresh 'pasta filata' cheese made from cow's milk and cream that is rapidly spreading in Europe. It has very high caloric content, and a technological protocol was developed for producing a reduced-fat type and fortifying it with polyunsaturated fatty acids (PUFA) of vegetable origin. A satisfactory reduced-fat prototype was obtained by using a 14% fat cream, which was specifically developed by diluting double cream with a suspension of carob seed flour. The composition of the new cheese changed with respect to the control, but the sensory characteristics were not impaired. Moisture increased from 62·6 to 68·4%, fat on dry matter decreased from 59·1 to 34·7%, and the caloric content decreased from 1060·8 to 718 J/100 g. Proteolysis and lipolysis were not affected by the technological modifications: after 7 d storage, the electrophoretic pattern of caseins and the free fatty acids profile of experimental and control cheeses were not significantly different. Fortification of reduced-fat Burrata with PUFA was obtained by using two commercial formulates available at a compatible price with the current economic values of the cheese. The two formulates derived from flaxseeds and Carthamus tinctorius oil and allowed enrichment in C18 :3 : n3 (α-linolenic acid, ALA), and 9cis,11trans- and 10trans,12cis- conjugated linoleic acid (CLA), respectively. Fortification was easy to perform under a technical point of view, but the negative sensory impact limited fortification at a maximum of 7·0 mg g-1 fat ALA and 6·8 g-1 fat CLA.

  1. Does Environmental Enrichment Reduce Stress? An Integrated Measure of Corticosterone from Feathers Provides a Novel Perspective

    Science.gov (United States)

    Fairhurst, Graham D.; Frey, Matthew D.; Reichert, James F.; Szelest, Izabela; Kelly, Debbie M.; Bortolotti, Gary R.

    2011-01-01

    Enrichment is widely used as tool for managing fearfulness, undesirable behaviors, and stress in captive animals, and for studying exploration and personality. Inconsistencies in previous studies of physiological and behavioral responses to enrichment led us to hypothesize that enrichment and its removal are stressful environmental changes to which the hormone corticosterone and fearfulness, activity, and exploration behaviors ought to be sensitive. We conducted two experiments with a captive population of wild-caught Clark's nutcrackers (Nucifraga columbiana) to assess responses to short- (10-d) and long-term (3-mo) enrichment, their removal, and the influence of novelty, within the same animal. Variation in an integrated measure of corticosterone from feathers, combined with video recordings of behaviors, suggests that how individuals perceive enrichment and its removal depends on the duration of exposure. Short- and long-term enrichment elicited different physiological responses, with the former acting as a stressor and birds exhibiting acclimation to the latter. Non-novel enrichment evoked the strongest corticosterone responses of all the treatments, suggesting that the second exposure to the same objects acted as a physiological cue, and that acclimation was overridden by negative past experience. Birds showed weak behavioral responses that were not related to corticosterone. By demonstrating that an integrated measure of glucocorticoid physiology varies significantly with changes to enrichment in the absence of agonistic interactions, our study sheds light on potential mechanisms driving physiological and behavioral responses to environmental change. PMID:21412426

  2. Opening address at the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    Kazuo Sato

    1984-01-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of variuos programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. More detailed status of the RERTR program in Japan, as the host country is covered in this presentation

  3. Can Halogen Enrichment in Reduced Enstatite Chondrites Provide Clues to Volatile Accretion in the Early Earth?

    Science.gov (United States)

    Clay, P. L.; Burgess, R.; Busemann, H.; Ruzié, L.; Joachim, B.; Ballentine, C.

    2013-12-01

    Understanding how the Earth obtained and ultimately retained its volatiles is important for our overall understanding of large scale planetary evolution. Numerous models exist for the heterogeneous accretion of volatiles to early Earth, but accounting for all elements through accretion of typical planetary building blocks (e.g., CI chondrites) is difficult. Proto-planetary collisions resulting in the accretion of volatile-poor material under reducing conditions followed by accretion of volatile-rich material under oxidizing conditions has been suggested in such models [e.g., 1]. The heavy halogens (Cl, Br and I), a group of moderately volatile elements, are excellent tracers of planetary processing due to their low abundance and incompatible nature. Therefore characterizing halogen abundance and distribution in materials that accreted to form the planets, e.g., primitive meteorites, is crucial. One group of primitive meteorites, the enstatite chondrites (EC's), are amongst the most reduced materials in the solar system as evidenced by their unique mineral assemblage. Yet despite forming under ultra-reducing conditions, they are enriched in the moderately volatile elements, such as the halogens. The ECs are of particular interest owing to their oxygen isotopic composition which plots along the terrestrial fractionation line, linking them isotopically to the Earth-Moon system. These samples can thus potentially provide clues on the accretion of moderately volatile element rich material under reducing conditions, such as it may have existed during the early stages of Earth's accretion. Chlorine, Br and I concentrations in ECs were determined through step-heating small neutron-irradiated samples (0.3 to 3.3 mg) and measured by mass spectrometry using the noble gas proxy isotopes 38ArCl/Cl, 80KrBr/Br and 128XeI/I. The EH chondrites are consistently enriched in the heavy halogens (up to 330 ppm Cl, 2290 ppb Br and 180 ppb I), compared to other ordinary and carbonaceous

  4. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  5. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

    The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

  6. Inerting of a Vented Aircraft Fuel Tank Test Article with Nitrogen-Enriched Air

    National Research Council Canada - National Science Library

    Burns, Michael

    2001-01-01

    ...) required to inert a vented aircraft fuel tank. NEA, generated by a hollow fiber membrane gas separation system, was used to inert a laboratory fuel tank with a single vent on top designed to simulate a transport category airplane fuel tank...

  7. 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)

    Zuhair; Suwoto

    2009-01-01

    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 k eff 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 k eff on fuel kernel radius decreases in relation to the increase in uranium enrichment . However, k eff 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)

  8. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

    This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

  9. Consideration of Nuclear Criticality When Directly Disposing Highly Enriched Spent Nuclear Fuel in Unsaturated Tuff - I: Nuclear Criticality Constraints

    International Nuclear Information System (INIS)

    Rechard, Rob P.; Sanchez, Lawrence C.; Trellue, Holly R.

    2003-01-01

    This paper presents the mass, concentration, and volume required for a critical event to occur in homogeneous mixtures of fissile material and various other geologic materials. The fissile material considered is primarily highly enriched uranium spent fuel; however, 239 Pu is considered in some cases. The non-fissile materials examined are those found in the proposed repository area at Yucca Mountain, Nevada: volcanic tuff, iron rust, concrete, and naturally occurring water. For 235 U, the minimum critical solid concentration for tuff was 5 kg/m 3 (similar to sandstone), and in goethite, 45 kg/m 3 . The critical mass of uranium was sensitive to a number of factors, such as moisture content and fissile enrichment, but had a minimum, assuming almost 100% saturation and >20% enrichment, of 18 kg in tuff as Soddyite (or 9.5 kg as UO 2 ) and 7 kg in goethite. For 239 Pu, the minimum critical solid concentration for tuff was 3 kg/m 3 (similar to sandstone); in goethite, 20 kg/m 3 . The critical mass of plutonium was also sensitive to a number of factors, but had a minimum, assuming 100% saturation and 80-90% enrichment, of 5 kg in tuff and 6 kg in goethite

  10. ALARA (As Low As Reasonable Achievable) procedure applied to fuel assembly fabrication with enriched reprocessing uranium (ERU)

    International Nuclear Information System (INIS)

    Guimaraes, Leonam dos Santos; Degrange, Jean Pierre

    1998-01-01

    The study introduced by this paper compose the first step to the implementation of ALARA (As Low As Reasonable Achievable) for a nuclear fuel assembly factory which one of its two production lines will be designed to work with Enriched Reprocessing Uranium (ERU). This step includes the reference situation analysis is based on previsional dosimetric evaluations for individual and collective exposures of each factory operator (117 in total) working on 7 work stations, considering 6 annual production scenarios (10, 50 75, 100 and 150 ERU tons), which corresponds to an annual production of 600 tons (ERU plus enriched natural uranium ENU). The exposure indicators evolution, expressed in terms of collective dose, annual individual dose and radiological detrimental cost for workers, is also used in a complimentary way to guide the analysis. (author)

  11. 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)

  12. Characterization of Fe (III)-reducing enrichment culture and isolation of Fe (III)-reducing bacterium Enterobacter sp. L6 from marine sediment.

    Science.gov (United States)

    Liu, Hongyan; Wang, Hongyu

    2016-07-01

    To enrich the Fe (III)-reducing bacteria, sludge from marine sediment was inoculated into the medium using Fe (OH)3 as the sole electron acceptor. Efficiency of Fe (III) reduction and composition of Fe (III)-reducing enrichment culture were analyzed. The results indicated that the Fe (III)-reducing enrichment culture with the dominant bacteria relating to Clostridium and Enterobacter sp. had high Fe (III) reduction of (2.73 ± 0.13) mmol/L-Fe (II). A new Fe (III)-reducing bacterium was isolated from the Fe (III)-reducing enrichment culture and identified as Enterobacter sp. L6 by 16S rRNA gene sequence analysis. The Fe (III)-reducing ability of strain L6 under different culture conditions was investigated. The results indicated that strain L6 had high Fe (III)-reducing activity using glucose and pyruvate as carbon sources. Strain L6 could reduce Fe (III) at the range of NaCl concentrations tested and had the highest Fe (III) reduction of (4.63 ± 0.27) mmol/L Fe (II) at the NaCl concentration of 4 g/L. This strain L6 could reduce Fe (III) with unique properties in adaptability to salt variation, which indicated that it can be used as a model organism to study Fe (III)-reducing activity isolated from marine environment. Copyright © 2015 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  13. Could reducing fossil-fuel emissions cause global warming

    Energy Technology Data Exchange (ETDEWEB)

    Wigley, T M.L. [University of East Anglia, Norwich (UK). Climatic Research Unit

    1991-02-07

    When fossil fuel is burned, both carbon dioxide and sulphur dioxide are added to the atmosphere. The former should cause warming of the lower atmosphere by enhancing the greenhouse effect, whereas the latter, by producing sulphate aerosols, may cause a cooling effect. The possibility that these two processes could offset each other was suggested many years ago but during most of the intervening period, attention has focused on the greenhouse effect. Interest in tropospheric aerosols has, however, recently been rekindled by the realization that they may influence climate, not only through clear-sky radiative effects, but also by modifying cloud albedo. The author examines the sensitivity of the climate system to simultaneous changes in SO{sub 2} and CO{sub 2} emissions, as might occur if controls were imposed on fossil-fuel use. Over the next 10-30 years, it is conceivable that the increased radiative forcing due to SO{sub 2} concentration changes could more than offset reductions in radiative forcing due to reduced CO{sub 2} emissions. 16 refs., 3 figs., 1 tab.

  14. Microbial fuel cell based on electroactive sulfate-reducing biofilm

    International Nuclear Information System (INIS)

    Angelov, Anatoliy; Bratkova, Svetlana; Loukanov, Alexandre

    2013-01-01

    Highlights: ► Regulation and management of electricity generation by variation of residence time. ► Design of microbial fuel cell based on electroactive biofilm on zeolite. ► Engineering solution for removing of the obtained elemental sulfur. - abstract: A two chambered laboratory scale microbial fuel cell (MFC) has been developed, based on natural sulfate-reducing bacterium consortium in electroactive biofilm on zeolite. The MFC utilizes potassium ferricyanide in the cathode chamber as an electron acceptor that derives electrons from the obtained in anode chamber H 2 S. The molecular oxygen is finally used as a terminal electron acceptor at cathode compartment. The generated power density was 0.68 W m −2 with current density of 3.2 A m −2 at 150 Ω electrode resistivity. The hydrogen sulfide itself is produced by microbial dissimilative sulfate reduction process by utilizing various organic substrates. Finally, elemental sulfur was identified as the predominant final oxidation product in the anode chamber. It was removed from MFC through medium circulation and gathering in an external tank. This report reveals dependence relationship between the progress of general electrochemical parameters and bacterial sulfate-reduction rate. The presented MFC design can be used for simultaneous sulfate purification of mining drainage wastewater and generation of renewable electricity

  15. ENVIRONMENTAL ENRICHMENT STRENGTHENS CORTICOCORTICAL INTERACTIONS AND REDUCES AMYLOID-β OLIGOMERS IN AGED MICE

    Directory of Open Access Journals (Sweden)

    Marco eMainardi

    2014-01-01

    Full Text Available Brain aging is characterized by global changes which are thought to underlie age-related cognitive decline. These include variations in brain activity and the progressive increase in the concentration of soluble amyloid-β (Aβ oligomers, directly impairing synaptic function and plasticity even in the absence of any neurodegenerative disorder. Considering the high social impact of the decline in brain performance associated to aging, there is an urgent need to better understand how it can be prevented or contrasted. Lifestyle components, such as social interaction, motor exercise and cognitive activity, are thought to modulate brain physiology and its susceptibility to age-related pathologies. However, the precise functional and molecular factors that respond to environmental stimuli and might mediate their protective action again pathological aging still need to be clearly identified. To address this issue, we exploited environmental enrichment (EE, a reliable model for studying the effect of experience on the brain based on the enhancement of cognitive, social and motor experience, in aged wild-type mice. We analyzed the functional consequences of EE on aged brain physiology by performing in vivo local field potential (LFP recordings with chronic implants. In addition, we also investigated changes induced by EE on molecular markers of neural plasticity and on the levels of soluble Aβ oligomers. We report that EE induced profound changes in the activity of the primary visual and auditory cortices and in their functional interaction. At the molecular level, EE enhanced plasticity by an upward shift of the cortical excitation/inhibition balance. In addition, EE reduced brain Aβ oligomers and increased synthesis of the Aβ-degrading enzyme neprilysin. Our findings strengthen the potential of EE procedures as a non-invasive paradigm for counteracting brain aging processes.

  16. Reducing the carbon footprint of fuels and petrochemicals. Preprints

    International Nuclear Information System (INIS)

    Ernst, S.; Balfanz, U.; Buchholz, S.; Lichtscheidl, J.; Marchionna, M.; Nees, F.; Santacesaria, E.

    2012-01-01

    Within the DGMK conference between 08th and 10th October, 2012, in Berlin (Federal Republic of Germany) the following lectures were held: (1) Energy demand and mix for global welfare and stable ecosystems (A. Jess); (2) The EU's roadmap for moving to a low-carbon economy - Aspirations and reality for refiners (J. Lichtscheidl); (3) Applications of CCS technology to the oil and gas industries (M. Marchionna); (4) A new chemical system solution for acid gas removal (M. Seiler); (5) Hydrogenation of carbon dioxide towards synthetic natural gas - A route to effective future energy storage (M. Schoder); (6) Bio-MTBE - How to reduce CO 2 footprint in fuels with a well known premium gasoline component (O. Busch); (7) Use of waste materials for Biodiesel production (R. Vitiello); (8) From algae to diesel and kerosene - Tailored fuels via selective catalysis (C. Zhao); (9) Chemo-catalytic valorization of cellulose (R. Palkovits); (10) Cellulosic ethanol: Potential, technology and development status (M. Rarbach); (11) Methanation of carbon oxides - History, status quo and future perspectives (W. Kaltner); (12) Chemical storage of renewable electricity in hydrocarbon fuels via H 2 (H. Eilers); (13) Materials for the 21st century: Can the carbon come from CO 2 (S. Kissling); (14) Effect of CO 2 admixture on the catalytic performance of Ni-Nb-M-O catalysts in oxidative dehydrogenation of ethane to ethylene (A. Qiao); (15) Oxidative dehydrogenation of light alkanes (A. Meiswinkel); (16) Low carbon fuel and chemical production from waste gases (S. Simpson); (17) Methanol to propylene: From development to commercialization (S. Haag); (18) On the impact of olefins and aromatics in the methanol-to-hydrocarbon conversion over H-ZSM-5 catalysts (X. Sun); (19) Mn-Na 2 WO 4 /SiO 2 - An industrial catalyst for methane coupling (M. Yildiz); (20) Biorefineries - Prerequisites for the realization of a future bioeconomy (K. Wagemann); (21) A new process for the valorisation of a bio

  17. Reducing the carbon footprint of fuels and petrochemicals. Preprints

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, S.; Balfanz, U.; Buchholz, S.; Lichtscheidl, J.; Marchionna, M.; Nees, F.; Santacesaria, E. (eds.)

    2012-07-01

    Within the DGMK conference between 08th and 10th October, 2012, in Berlin (Federal Republic of Germany) the following lectures were held: (1) Energy demand and mix for global welfare and stable ecosystems (A. Jess); (2) The EU's roadmap for moving to a low-carbon economy - Aspirations and reality for refiners (J. Lichtscheidl); (3) Applications of CCS technology to the oil and gas industries (M. Marchionna); (4) A new chemical system solution for acid gas removal (M. Seiler); (5) Hydrogenation of carbon dioxide towards synthetic natural gas - A route to effective future energy storage (M. Schoder); (6) Bio-MTBE - How to reduce CO{sub 2} footprint in fuels with a well known premium gasoline component (O. Busch); (7) Use of waste materials for Biodiesel production (R. Vitiello); (8) From algae to diesel and kerosene - Tailored fuels via selective catalysis (C. Zhao); (9) Chemo-catalytic valorization of cellulose (R. Palkovits); (10) Cellulosic ethanol: Potential, technology and development status (M. Rarbach); (11) Methanation of carbon oxides - History, status quo and future perspectives (W. Kaltner); (12) Chemical storage of renewable electricity in hydrocarbon fuels via H{sub 2} (H. Eilers); (13) Materials for the 21st century: Can the carbon come from CO{sub 2} (S. Kissling); (14) Effect of CO{sub 2} admixture on the catalytic performance of Ni-Nb-M-O catalysts in oxidative dehydrogenation of ethane to ethylene (A. Qiao); (15) Oxidative dehydrogenation of light alkanes (A. Meiswinkel); (16) Low carbon fuel and chemical production from waste gases (S. Simpson); (17) Methanol to propylene: From development to commercialization (S. Haag); (18) On the impact of olefins and aromatics in the methanol-to-hydrocarbon conversion over H-ZSM-5 catalysts (X. Sun); (19) Mn-Na{sub 2}WO{sub 4}/SiO{sub 2} - An industrial catalyst for methane coupling (M. Yildiz); (20) Biorefineries - Prerequisites for the realization of a future bioeconomy (K. Wagemann); (21) A new process

  18. Preliminary study for the transport of the fuel rods of U235 enriched to 1.8 per cent

    International Nuclear Information System (INIS)

    Cardenas, H.; Perez, A.

    1998-01-01

    Transport of 1,8% U235 enriched fuel rods needs both the evaluation of the radiological risk and considerations about criticality aspects. Issues as diverse production characteristics, storage facilities in the source of origin an economical aspects have to be added to the radiological and nuclear considerations. Transport of those rods through national territory must comply with the Argentine Regulatory authority's regulations, based on the Safety Series No. 6, (ed. 1985) -as amended 1990- IAEA. Safety criteria are exposed, taking into account the amount of material to be transported, container characteristics, packaging type and expedition conditions. (author)

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

    International Nuclear Information System (INIS)

    Arbie, B.; Soentono, S.

    1987-01-01

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

  20. Progress in development of low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

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

    2002-01-01

    Results from post irradiation examinations and analyses of U-Mo/Al dispersion mini plates are presented. Irradiation test RERTR-5 contained mini- fuel plates with fuel loadings of 6 and 8 g U cm -3 . The fuel material consisted of 6, 7 and 10 wt. % Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion. (author)

  1. A comparison of integral block and tubular interacting fuel element concepts for low enrichment HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A

    1972-04-15

    The tubular interacting fuel element has to date been the favoured U.K. high temperature reactor design. Recent attempts to lower fuel costs and the progress of the Fort St. Vrain reactor has focussed attention on alternative designs, and in particular on the attractive design simplicity of the integral block concept. The aim of this investigation is to compare the merits of both concepts from fuel cycle cost and thermal performance viewpoints and to determine whether optimization of the integral block concept leads to changes in the current design values of (a) fuel density, (b) Nc/Nu, and/or (c) mean discharge irradiation within the framework of present design limits.

  2. 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.

  3. Allocation of uranium enrichment services to fuel foreign and domestic nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    This interim report was made in response to a request for information concerning the sale of U.S. uranium enrichment services to foreign countries and its effect on AEC's ability to meet domestic demands. Long-term enrichment services (June 30, 1974), both domestic and foreign, totaled 364,000 MW, or 44,000 MW more than its available capability. The first-come-first-served policy was modified to give preferential treatment to Yugoslav and Mexican requests because of IAEA commitments, and to shift six standard contracts from Japan. From Aug. to Sept. 1974, standard contracts were signed for all 15 pending domestic requests and for 33 pending foreign requests, with the remaining 45 foreign requests depending on NRC's approval of Pu recycle, although private enrichment or stockpile enriched U could meet these needs. There is no firm commitment in the private sector to build and operate the needed enrichment plant. The acceleration of foreign nuclear programs coupled with ERDA's termination of further long-term contracts, may lead to the emergence of foreign supply sources, and U.S. may lose its favorable balance-of-payments and its influence on international nuclear policies

  4. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  5. Summary report on the aerobic degradation of diesel fuel and the degradation of toluene under aerobic, denitrifying and sulfate reducing conditions

    International Nuclear Information System (INIS)

    Coyne, P.; Smith, G.

    1995-01-01

    This report contains a number of studies that were performed to better understand the technology of the biodegradation of petroleum hydrocarbons. Topics of investigation include the following: diesel fuel degradation by Rhodococcus erythropolis; BTEX degradation by soil isolates; aerobic degradation of diesel fuel-respirometry; aerobic degradation of diesel fuel-shake culture; aerobic toluene degradation by A3; effect of HEPES, B1, and myo-inositol addition on the growth of A3; aerobic and anaerobic toluene degradation by contaminated soils; denitrifying bacteria MPNs; sulfate-reducing bacteria MPNs; and aerobic, DNB and SRB enrichments

  6. Anaerobic degradation of propane and butane by sulfate-reducing bacteria enriched from marine hydrocarbon cold seeps.

    Science.gov (United States)

    Jaekel, Ulrike; Musat, Niculina; Adam, Birgit; Kuypers, Marcel; Grundmann, Olav; Musat, Florin

    2013-05-01

    The short-chain, non-methane hydrocarbons propane and butane can contribute significantly to the carbon and sulfur cycles in marine environments affected by oil or natural gas seepage. In the present study, we enriched and identified novel propane and butane-degrading sulfate reducers from marine oil and gas cold seeps in the Gulf of Mexico and Hydrate Ridge. The enrichment cultures obtained were able to degrade simultaneously propane and butane, but not other gaseous alkanes. They were cold-adapted, showing highest sulfate-reduction rates between 16 and 20 °C. Analysis of 16S rRNA gene libraries, followed by whole-cell hybridizations with sequence-specific oligonucleotide probes showed that each enrichment culture was dominated by a unique phylotype affiliated with the Desulfosarcina-Desulfococcus cluster within the Deltaproteobacteria. These phylotypes formed a distinct phylogenetic cluster of propane and butane degraders, including sequences from environments associated with hydrocarbon seeps. Incubations with (13)C-labeled substrates, hybridizations with sequence-specific probes and nanoSIMS analyses showed that cells of the dominant phylotypes were the first to become enriched in (13)C, demonstrating that they were directly involved in hydrocarbon degradation. Furthermore, using the nanoSIMS data, carbon assimilation rates were calculated for the dominant cells in each enrichment culture.

  7. Results of Cesar II critical facility with low enriched fuel balls

    Energy Technology Data Exchange (ETDEWEB)

    Langlet, G; Guerange, J; Laponche, B; Morier, F; Neef, R D; Bock, H J; Kring, F J; Scherer, W

    1972-06-15

    The Cesar facility has been transformed to load in its center a pebble bed fuel. This new Cesar assembly is called Cesar II. The program for the measurements with HTR type fuel balls is managed under a cooperation between physicists of CEA/CADARACHE and KFA/JUELICH. A description of the measuring zones of Cesar II and of the experimental results is given.

  8. The Potential of Turboprops to Reduce Aviation Fuel Consumption

    OpenAIRE

    Smirti, Megan; Hansen, Mark

    2009-01-01

    Aviation system planning, particularly fleet selection and adoption, is challenged by fuel price uncertainty. Fuel price uncertainty is due fuel and energy price fluctuations and a growing awareness of the environmental externalities related to transportation activities, particularly as they relate to climate change. To assist in aviation systems planning under such fuel price uncertainty and environmental regulation, this study takes a total logistic cost approach and evaluates three represe...

  9. Optimization to reduce fuel consumption in charge depleting mode

    Science.gov (United States)

    Roos, Bryan Nathaniel; Martini, Ryan D.

    2014-08-26

    A powertrain includes an internal combustion engine, a motor utilizing electrical energy from an energy storage device, and a plug-in connection. A Method for controlling the powertrain includes monitoring a fuel cut mode, ceasing a fuel flow to the engine based upon the fuel cut mode, and through a period of operation including acceleration of the powertrain, providing an entirety of propelling torque to the powertrain with the electrical energy from the energy storage device based upon the fuel cut mode.

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

    International Nuclear Information System (INIS)

    Almeida, M.C.M. de.

    1988-01-01

    The enrichment of 235 U in UO 2 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 235 U and 239 Np different energies peaks from 238 U, 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.) [pt

  11. Sogin enriched uranium extraction (EUREX) plant spent fuel pool cleaning and decontamination utilizing the Smart Safe solution

    International Nuclear Information System (INIS)

    Denton, M.S.; Gili, M.; Nasta, M.; Quintiliani, R.; Caccia, G.; Botzen, W.; Forrester, K.

    2009-01-01

    SOGIN's EUREX facility in Italy was developed as a pilot plant functional testing laboratory for spent fuel reprocessing. This facility was operated successfully for many years since 1970 and was eventually shutdown consistent with Italy's suspension of all nuclear operations. At the time of suspension, the EUREX facility still had spent nuclear fuel assemblies in storage from a nearby PWR. Other fuel assemblies from an Italian AGR had remained stored in the spent fuel pool for the 20 years or so waiting for removal and reprocessing abroad. Being Magnox fuel elements, their recovery for the transport produced a huge amount of sludge in the pool. During this time, sediment, dirt, corrosion products, fuel cladding, etc. has collected within the fuel pool as a crud layer dispersed throughout. Most of this crud has accumulated on the horizontal surfaces of the pool and fuel element assemblies, while some remains as a suspended colloidal material. Furthermore many other contaminated metal components, used during the operation years, where still inside the pool. Due to a pool leak discovered in 2006, SOGIN speeded up its pool decommissioning program, making also available the transfer of the spent fuel to a nearby interim repository, with the goal to completely drain the pool in the shortest period of time. In order for SOGIN to successfully transfer the fuel assemblies from their current storage basket locations to the spent fuel transfer cask, the bulk of the crud needed to be removed. This cleanup operation was deemed necessary to minimize the suspension of contamination in the water during underwater handling operations. This would reduce the decontamination efforts on the transfer cask upon removal, once loaded with the spent fuel, and enhance safety by reducing potential underwater visibility issues. The operations were completed in July 2008 with the release to the environment of the pool water, thoroughly purified and without any relevant radiological impact. The

  12. Nuclear fuel cycle, nuclear fuel makes the rounds: choosing a closed fuel cycle, nuclear fuel cycle processes, front-end of the fuel cycle: from crude ore to enriched uranium, back-end of the fuel cycle: the second life of nuclear fuel, and tomorrow: multiple recycling while generating increasingly less waste

    International Nuclear Information System (INIS)

    Philippon, Patrick

    2016-01-01

    France has opted for a policy of processing and recycling spent fuel. This option has already been deployed commercially since the 1990's, but will reach its full potential with the fourth generation. The CEA developed the processes in use today, and is pursuing research to improve, extend, and adapt these technologies to tomorrow's challenges. France has opted for a 'closed cycle' to recycle the reusable materials in spent fuel (uranium and plutonium) and optimise ultimate waste management. France has opted for a 'closed' nuclear fuel cycle. Spent fuel is processed to recover the reusable materials: uranium and plutonium. The remaining components (fission products and minor actinides) are the ultimate waste. This info-graphic shows the main steps in the fuel cycle currently implemented commercially in France. From the mine to the reactor, a vast industrial system ensures the conversion of uranium contained in the ore to obtain uranium oxide (UOX) fuel pellets. Selective extraction, purification, enrichment - key scientific and technical challenges for the teams in the Nuclear Energy Division (DEN). The back-end stages of the fuel cycle for recycling the reusable materials in spent fuel and conditioning the final waste-forms have reached maturity. CEA teams are pursuing their research in support of industry to optimise these processes. Multi-recycle plutonium, make even better use of uranium resources and, over the longer term, explore the possibility of transmuting the most highly radioactive waste: these are the challenges facing future nuclear systems. (authors)

  13. Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

    International Nuclear Information System (INIS)

    Sarmiento, G.S.

    1980-01-01

    Within natural UO 2 fuel elements enriched with plutonium, this last material should form PuO 2 solid solutions inside the UO 2 pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO 2 islands are formed in the UO 2 matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium. A detailed study of the thermal effects produced by PuO 2 particles in the UO 2 matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spatial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a well-known exact solution, has been obtained. (orig.)

  14. The economic impact of strengthening fuel quality regulation-reducing sulfur content in diesel fuel

    International Nuclear Information System (INIS)

    Chang, H.J.; Cho, G.L.; Kim, Y.D.

    2006-01-01

    This paper investigates the impact of strengthening vehicle emission regulation on economic activities. The government attempts to use three regulation measures to protect air quality from transportation emission. The measures include the aggregate limit (bubbles), the vehicle emission standard, and the fuel quality standard. Especially, we focus on the economic impact of reducing sulfur content in diesel fuel quality standard. Sulfur content in diesel fuel is one of the main factors in worsening local air quality. The emission from diesel vehicle accounts for 51.8% of total vehicle emission in Korea. If sulfur content reduction regulation is implemented, then the petroleum industry should build more facility to produce low sulfur content diesel, leading to additional production costs and increasing prices and decreasing outputs. We use computable general equilibrium model to analyze how the sulfur reduction regulation affects economic activities and trace out local emission reduction cost and GDP loss. And we suggest the tax-recycling mechanism to mitigate the negative economic costs due to the sulfur reduction regulation

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  16. Operational experience with the first eighteen slightly enriched uranium fuel assemblies in the Atucha-1 nuclear power plant

    International Nuclear Information System (INIS)

    Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Fink, J.; Casario, J.A.; Alvarez, L.

    1997-01-01

    Atucha I is a 357 Mwe nuclear station, moderated and cooled with heavy water, pressure vessel type of German design, located in Argentina. Fuel assemblies (FA) are 36 active natural UO2 rod clusters, 5.3 meters long and fuel is on power. Average FA exit burnup is 6 MWd/kg U. The reactor core contains 252 FA. To reduce the fuel costs about 6 MU$S/yr a program of utilization of SEU (0.85 %w U235) fuel was started at the beginning of 1995 with the introduction of 12 FA in the first step. The exit burnup of FA is approx. 10 MWd/kgU. It is planned to increase gradually the number of them up to having a full core with SEU fuel with an expected FA average exit burnup of 11 MWd/kgU. The SEU program has also the advantage of a strong reduction of spent fuel volume, and a moderate reduction of fuelling machine use. This paper presents the satisfactory operation experience with the introduction of the first 12 SEU fuel assemblies and the planned activities for the future. The fresh SEU fuel assemblies were introduced in six fuel channels located in an intermediate zone located 136 cm from the center of the reactor and selected because they have higher margins to the channel powers limits to accommodate the initial 15 to 20 % relative channel power increase. To verify the design and fuel management calculations, comparisons have been made of the calculated and measured values of the variation of channel ΔT, regulating rods insertion and flux reading in in-core detectors near to the refueled channel. The agreement was good and in most of the cases within the measurement errors. Cell calculations were made with WIMS-D4, and reactor calculations with PUMA. a fuel management 3D diffusion program developed in Argentina. With SEU fuel with a greater burnup in the central high power core region, new operating procedures were developed to prevent PCI failures in fuel power ramps that arise during operation. Some fuel rod and structural assembly design changes were introduced on the

  17. Physical and welding metallurgy of Gd-enriched austenitic alloys for spent nuclear fuel applications. Part II, nickel base alloys

    International Nuclear Information System (INIS)

    Mizia, Ronald E.; Michael, Joseph Richard; Williams, David Brian; Dupont, John Neuman; Robino, Charles Victor

    2004-01-01

    The physical and welding a metallurgy of gadolinium- (Gd-) enriched Ni-based alloys has been examined using a combination of differential thermal analysis, hot ductility testing. Varestraint testing, and various microstructural characterization techniques. Three different matrix compositions were chosen that were similar to commercial Ni-Cr-Mo base alloys (UNS N06455, N06022, and N06059). A ternary Ni-Cr-Gd alloy was also examined. The Gd level of each alloy was ∼2 wt-%. All the alloys initiated solidification by formation of primary austenite and terminated solidification by a Liquid γ + Ni 5 Gd eutectic-type reaction at ∼1270 C. The solidification temperature ranges of the alloys varied from ∼100 to 130 C (depending on alloy composition). This is a substantial reduction compared to the solidification temperature range to Gd-enriched stainless steels (360 to 400 C) that terminate solidification by a peritectic reaction at ∼1060 C. The higher-temperature eutectic reaction that occurs in the Ni-based alloys is accompanied by significant improvements in hot ductility and solidification cracking resistance. The results of this research demonstrate that Gd-enriched Ni-based alloys are excellent candidate materials for nuclear criticality control in spent nuclear fuel storage applications that require production and fabrication of large amounts of material through conventional ingot metallurgy and fusion welding techniques

  18. The low enriched fuel cycle in the GA 1160 MW design and the switch-over to thorium

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, H.

    1974-03-15

    Calculations for the GA 1160 MW HTR are presented. The aim of these investigations was to compare the Low Enriched Uranium (LEU) cycle and the Thorium cycle for the GA 1160 MW HTR both using the same GA designed integral block fuel element. The total fuel cycle cost for the equilibrium cycle comes out to be about 16% cheaper for the Thorium cycle than for the Low-Enriched cycle. However, these favorable results for the thorium cycle are completely dependent on the availability of reprocessing and refabrication facilities, for costs comparable with the costs used for these investigations. The possibility of starting the reactor on a LEU 3 year cycle and later switching over to a thorium 4 year cycle was investigated. No cost penalties were found to be paid during the switch-over. The problems of local power peaks and age factors were not investigated in greater detail as only integral physical quantities were obtained from the neutron physics calculations. However, no indications of any problem in the switch-over phase were given. Elaborate 3-dimensional methods are necessary for further investigation of these types of problems.

  19. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)

    2008-09-15

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

  20. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  1. Nutrient enrichment reduces constraints on material flows in a detritus-based food web

    Science.gov (United States)

    Wyatt F. Cross; Bruce Wallace; Amy D. Rosemond

    2007-01-01

    Most aquatic and terrestrial ecosystems are experiencing increased nutrient availability, which is affecting their structure and function. By altering community composition and productivity of consumers, enrichment can indirectly cause changes in the pathways and magnitude of material flows in food webs. These changes, in turn, have major consequences for material...

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

    International Nuclear Information System (INIS)

    Taboada, Horacio; Gonzalez, Alfredo G.

    2005-01-01

    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) [es

  3. Development program for fuel elements with low enriched uranium for high temperature reactors

    International Nuclear Information System (INIS)

    1987-12-01

    The results of HTR fuel development taking place at the THTR's can be summarized as follows for the main points of core manufacture coating matrix and fuel emenent manufacture: 1. The well known gel precipitation process was modified for the manufacture of UO 2 cores. 2. The TRISO coating (additional SiC layer between two very dense PyC layers) can be applied with the required quality on an economical 10 kg scale. 3. The particle fracture in the complete fuel element due to manufacture was lowered during the course of the project to below the target values of -6 U/U total. For testing fuel elements, the required irradiation samples were designed in agreement with the reactor constructors, were prepared and the first phase of the irradiation program was successfully completed in the context of the HBK project. (orig./HP) [de

  4. Pyrochemical recovery of easily reducible species from spent nuclear fuel

    International Nuclear Information System (INIS)

    Jouault, C.

    2000-01-01

    The purpose of the reprocessing of spent fuel is to separate noble metals and other easily reducible species, actinides and lanthanides. A thermodynamic and bibliographical study allowed us to elaborate a process which realises these separations in several steps. The experimental validation of the steps concerning the extraction of noble metals and easily reducible species required to imagine an apparatus which is conformed to the study of the two steps in question: the reduction by a gas of fission product oxides and the extraction of the metallic particles, obtained by reduction, by digestion in a liquid metal. Experiments on digestion, carried on molybdenum and ruthenium particles, allowed us to conclude that the transfer of metallic particles from a molten salt into a liquid metal is ruled by phenomena of complex wettability between the metallic particle, the molten salt, the liquid metal and the gas. The transfer from the salt to the metal is a chain of two steps: emersion of the particles from the salt to go into the gas, and then transfer from the gas into the metal. Kinetics are limited by the transfer through the metal surface. Kinetics study withdrew the experimental parameters and the metals properties which influence the digestion rate. A model on the transfer into a liquid metal of a particle trapped at the fluid/metal interface ratified the experimental conclusions and informed on the stirring influence. All the results allow us to think that the extraction of noble metals and easily reducible species are feasible in this way. (author) [fr

  5. Oxygen enrichment incineration

    International Nuclear Information System (INIS)

    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

  6. 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.

  7. 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

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

    Science.gov (United States)

    Zanaroli, Giulio; Di Toro, Sara; Todaro, Daniela; Varese, Giovanna C; Bertolotto, Antonio; Fava, Fabio

    2010-02-16

    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. 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 degrees 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 degrees C for several months. ENZ-G1 and ENZ-G2 are very similar highly enriched consortia of bacteria and a fungus capable of

  9. Critical experiments simulating accidental water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Glushkov, L.S.

    2003-01-01

    The paper focuses on experimental analysis of nuclear criticality safety at accidental water immersion of fuel elements of the Russian TOPAZ-2 space nuclear power system reactor. The structure of water-moderated heterogeneous critical assemblies at the NARCISS facility is described in detail, including sizes, compositions, densities of materials of the main assembly components for various core configurations. Critical parameters of the assemblies measured for varying number of fuel elements, height of fuel material in fuel elements and their arrangement in the water moderator with a uniform or variable spacing are presented. It has been found from the experiments that at accidental water immersion of fuel elements involved, the minimum critical mass equal to approximately 20 kg of uranium dioxide is achieved at 31-37 fuel elements. The paper gives an example of a physical model of the water-moderated heterogeneous critical assembly with a detailed characterization of its main components that can be used for calculations using different neutronic codes, including Monte Carlo ones. (author)

  10. Advances in uranium enrichment processes

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.; Slater, J.B.

    1986-05-01

    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

  11. 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

  12. Fabrication and irradiation testing of LEU [low enriched uranium] fuels at CRNL status as of 1987 September

    International Nuclear Information System (INIS)

    Sears, D.F.; Berthiaume, L.C.; Herbert, L.N.

    1987-01-01

    The current status of Chalk River Nuclear Laboratories' (CRNL) program to develop and test low-enriched uranium (LEU), proliferation-resistant fuels for use in research reactors is reviewed. CRNL's fuel manufacturing process has been qualified by the successful demonstration irradiation of 7 full-size rods in the NRU reactor. Now industrial-scale production equipment has been commissioned, and a fuel-fabrication campaign for 30 NRU rods and a MAPLE-X core is underway. Excess capacity could be used for commercial fuel fabrication. In the irradiation testing program, mini-elements with deliberately included core surface defects performed well in-reactor, swelling by only 7 to 8 vol% at 93 atomic percent burnup of the original U-235. The additional restraint provided by the aluminium cladding which flowed into the defects during extrusion contributed to this good performance. Mini-elements containing a variety of particle size distributions were also successfully irradiated to 93 at% burnup in NRU, as part of a study to establish the optimum particle size distribution. Swelling was found to be proportional to the percentage of fines (<44μm particles) contained in the cores. The mini-elements containing the composition normally used at CRNL had swollen by 5.8 vol%, and mini-elements with a much higher percentage of fines had swollen by 6.8 vol%, at 93 at% burnup. Also, a program to develop LEU targets for Mo-99 production, via the technology developed to fabricate dispersed silicide fuel, has started, and preliminary scoping studies are underway. (Author)

  13. Description of the CNEA U308 powder production plant for low enrichment fuel plates

    International Nuclear Information System (INIS)

    Boero, N.L.; Celora, J.; Parodi, C.A.; Pertossi, F.R.; Marajofsky, A.

    1987-01-01

    The design of the 20% enriched U 3 O 8 powder production plant was based on laboratory level experiments. The UF 6 hydrolysis, ADU precipitation, U 3 O 8 conversion processes were used. The equipment, controls and confinement were set not only by the processes but also by safety requirements according to the kind and physical form of the uranium compounds in each stage and criticality considerations. This paper describes the installation, set up and operation of the plant during production. (Author)

  14. Control concepts for vehicle drive line to reduce fuel consumption

    Energy Technology Data Exchange (ETDEWEB)

    Ossyra, J.C.

    2005-07-01

    In this work advanced drive line control concepts for off-road vehicles have been developed and investigated to reduce the power losses and finally the fuel consumption of the entire drive system by use of on-line optimization procedure. Two separate closed loop speed controls have been developed for the use on a microcontroller onboard the vehicle: one to control the hydrostatic transmission and the other to control the engine speed. Considering the loss characteristics of the displacement machines in the hydrostatic transmission and the steady state characteristics of the combustion engine by use of pure mathematical approximations of measured curves, a direct optimization strategy is used, which works on-line on a microcontroller. A laboratory hardware-in-the loop test rig has been used to investigate the proposed control concepts. For different typical and desired work cycles of an off-road machine on level ground and uphill a slope the effectiveness of the proposed control concepts have been proven. (orig.)

  15. Progress in developing very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

    1999-01-01

    Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

  16. Cooperative efforts for the removal of high-enriched fresh fuel from the Vinca Institute of Nuclear Sciences

    International Nuclear Information System (INIS)

    Hopwood, W.; Moses, S.; Pesic, M.; Sotic, O.; Wander, T.

    2003-01-01

    In August 2002, the inventory of high-enriched uranium (HEU) fresh fuel at the Vinca Institute in Belgrade, Yugoslavia, was repackaged and shipped to the Russian Federation (R.F.), its country of origin under the former Soviet Union. Several thousand small fuel elements were repackaged by the Vinca Institute into approved shipping containers provided by the RF and loaded onto the approved ground transportation vehicle. The transportation from the Vinca Institute to the Belgrade Airport was done under the planning and protection of Yugoslavian and Serbian military and police organizations, with technical oversight being provided by the Vinca staff that escorted the convoy. Under constant security protection, the Russian crew loaded the fuel containers onto the cargo plane, and later it departed for an airport near Dimitrovgrad, Russia. In addition to the domestic control and accounting provided during this operation, this inventory was under International Atomic Energy Agency (IAEA) safeguards, and its inspectors appropriately confirmed, sealed and documented the inventory. The United States (U.S.) observers were also present, and appropriate data were collected because of nonproliferation interests and contractual support for all phases of the operation. Since this event, the Vinca staff has generated several papers describing the technical background and detailed activities of this operation. This paper describes the removal from the U.S. observers perspectives and recognizes the significant cooperation among the supporting countries and the achievements of the organizations directly involved. (author)

  17. 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)

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    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

  18. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    International Nuclear Information System (INIS)

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-01-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  19. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  20. Methods for reducing lipid oxidation in fish-oil-enriched energy bars

    DEFF Research Database (Denmark)

    Nielsen, Nina Skall; Jacobsen, Charlotte

    2009-01-01

    P>Fish oil (FO) enrichment of foods is relevant owing to the beneficial effects of omega-3 polyunsaturated fatty acids on human health. However, the susceptibility of FO to oxidation necessitates careful control to avoid this oxidation. In this study, energy bars were successfully supplemented...... similar protection towards oxidation as packaging the energy bars in modified atmosphere. These protection methods were although not as efficient as addition of FO as micro-encapsulated powder. Addition of the metal chelator ethylene diamine tetra-acetic acid (EDTA) (100-2000 ppm) to emulsified FO...

  1. 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

  2. Evaluation of neutronic characteristics of STACY 80-cm-diameter cylindrical core fueled with 6% enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Sono, Hiroki

    2003-06-01

    For the examination of neutronic safety design of forthcoming experimental core configurations in the Static Experiment Critical Facility (STACY), neutronic characteristics of 80-cm-diameter cylindrical cores fueled with 6% enriched uranyl nitrate solution have been evaluated by computational analyses. In the analyses, the latest nuclear data library, JENDL-3.3, was used as neutron cross section data. The neutron diffusion and transport calculations were performed using a diffusion code, CITATION, in the SRAC code system and a continuous-energy Monte Carlo code, MVP. Critical level heights of the cores were obtained using such parameters as uranium concentration (up to 500 gU/l), free nitric acid concentration (up to 8 mol/l), and concentration of soluble neutron poisons, gadolinium and boron. It has been confirmed from the evaluation that all critical cores comply with safety criteria required in the STACY operation concerning excess reactivity, reactivity addition rates and shutdown margins by safety rods. (author)

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

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

    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

  4. Alternative Fuels in Transportation : Workforce needs and opportunities in support of reducing reliance on petroleum fuels

    Science.gov (United States)

    2016-01-01

    An overreliance on foreign oil and the negative impacts of using petroleum fuels on the worlds climate have prompted energy policies that support the diversification of transport fuels and aggressive work to transition to non-petroleum options. Th...

  5. Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

    2003-01-01

    In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

  6. Reducing DoD Fossil-Fuel Dependence

    Science.gov (United States)

    2006-09-01

    domestic market for demand and consumption of fossil fuel alternatives, or to drive fuel and transportation technology developments, in general. Barring...wholesale to the power market . IPPs own and operate their stations as non-utilities and do not own the transmission lines. Joule The (kinetic) energy acquired...maturiry for its seed. [Wikipedia, 13Aug06] TW Terawatt = 1012 Watts UAV Unmanned/Unpiloted Air Vehicle UCG Underground coal gasification USDA U.S

  7. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  8. Method for processing coal-enrichment waste with solid and volatile fuel inclusions

    Science.gov (United States)

    Khasanova, A. V.; Zhirgalova, T. B.; Osintsev, K. V.

    2017-10-01

    The method relates to the field of industrial heat and power engineering. It can be used in coal preparation plants for processing coal waste. This new way is realized to produce a loose ash residue directed to the production of silicate products and fuel gas in rotary kilns. The proposed method is associated with industrial processing of brown coal beneficiation waste. Waste is obtained by flotation separation of rock particles up to 13 mm in size from coal particles. They have in their composition both solid and volatile fuel inclusions (components). Due to the high humidity and significant rock content, low heat of combustion, these wastes are not used on energy boilers, they are stored in dumps polluting the environment.

  9. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    Science.gov (United States)

    Gorman, Phillip Michael

    possible to increase the amount of boron in the control blades by changing the assembly and core design. Nonetheless, the uncertainties in the multiplication factor due to nuclear data and void fraction uncertainty were assessed for the RBWR-SSH and the RBWR-TR, as well as for the RBWR-TB2. In addition, the uncertainty associated with the change in reactor states (such as the reactivity insertion in flooding the core) due to nuclear data uncertainties was quantified. The thorium RBWRs have much larger uncertainty of their DU-fueled counterparts as designed by Hitachi, as the fission cross section of 233U has very large uncertainty in the epithermal energy range. The uncertainty in the multiplication factor at reference conditions was about 1350 pcm for the RBWR-SSH, while it was about 900 pcm for the RBWR-TR. The uncertainty in the void coefficient of reactivity for both reactors is between 8 and 10 pcm/% void, which is on the same order of magnitude as the full core value. Finally, since sharp linear heat rate spikes were observed in the RBWR-TB2 simulation, the RBWR-TB2 unit cell was simulated using a much finer mesh than is possible using deterministic codes. It was found that the thermal neutrons reflecting back from the reflectors and the blankets were causing extreme spikes in the power density near the axial boundaries of the seeds, which were artificially smoothed out when using coarser meshes. It is anticipated that these spikes will cause melting in both seeds in the RBWR-TB2, unless design changes--such as reducing the enrichment level near the axial boundaries of the seeds--are made.

  10. The use of low enriched uranium fuel cycle in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    The present paper begins with a brief review of the status of research and development of experimental VHTR in Japan. On the basis of the experience gained from these work, assessment is made of commercial HTRs. Material balance with fuel burnup is calculated for the two core models; one is HTGR for steam cycle and the other VHTR for process heat application. The results of assessment of commercial HTRs are compared with those for LWR

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

    International Nuclear Information System (INIS)

    Castillo, Alejandro; Martín-del-Campo, Cecilia; Montes-Tadeo, José-Luis; François, Juan-Luis; Ortiz-Servin, Juan-José; Perusquía-del-Cueto, Raúl

    2014-01-01

    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

  12. 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)

    Eccleston, G.W.; Schrandt, R.G.; MacDonald, J.L.; Cverna, F.H.

    1979-01-01

    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 235 U. Material accountability measurements of the waste must be obtained in the presence of large neutron (0.5 x 10 6 n/s) and gamma (50,000 R/hr) backgrounds. The assay system employs fast-neutron irradiation of the sample, using a 5 mg 252 Cf 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 235 U is achievable with an end-of-life (1 mg) 252 Cf source and a background rate of 0.5 x 10 6 n/s

  13. Development of very high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-02-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  14. Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

    2002-01-01

    Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis

  15. Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor

    International Nuclear Information System (INIS)

    Jujuratisbela, U; Arbie, B; Pinem, S.; Tukiran; Suparlina, L.; Singh, O.P.

    1995-01-01

    Kinetics parameter measurements, such as reactivity worths of control rods and fuel elements, beam tube void reactivity, power reactivity coefficient and xenon poisoning reactivity have been performed on different cores of Reaktor Serba Guna G.A. Siwabessy (RSG-GAS). In parallel, a programme was also initiated to measure the other kinetics parameters like effective delayed neutron life time, prompt neutron decay constant, validation of period reactivity relationship and zero power frequency response function. The paper provides the results of these measurements. (author)

  16. Landing on empty: estimating the benefits from reducing fuel uplift in US Civil Aviation

    International Nuclear Information System (INIS)

    Ryerson, Megan S; Hansen, Mark; Hao, Lu; Seelhorst, Michael

    2015-01-01

    Airlines and Air Navigation Service Providers are united in their goal to reduce fuel consumption. While changes to flight operations and technology investments are the focus of a number of studies, our study is among the first to investigate an untapped source of aviation fuel consumption: excess contingency fuel loading. Given the downside risk of fuel exhaustion of diverting to an alternate airport, airline dispatchers may load excess fuel onto an aircraft. Such conservatism comes at a cost of consuming excess fuel, as fuel consumed is a function of, among other factors, aircraft weight. The aim of this paper is to quantify, on a per-flight basis, the fuel burned due to carrying fuel beyond what is needed for foreseeable contingencies, and thereby motivate research, federal guidance, and investments that allow airline dispatchers to reduce fuel uplift while maintaining near zero risks of fuel exhaustion. We merge large publicly available aviation and weather databases with a detailed dataset from a major US airline. Upon estimating factors that capture the quantity fuel consumed due to carrying a pound of weight for a range of aircraft types, we calculate the cost and greenhouse gas emissions from carrying unused fuel on arrival and additional contingency fuel above a conservative buffer for foreseeable contingencies. We establish that the major US carrier does indeed load fuel conservatively. We find that 4.48% of the fuel consumed by an average flight is due to carrying unused fuel and 1.04% of the fuel consumed by an average flight is due to carrying additional contingency fuel above a reasonable buffer. We find that simple changes in flight dispatching that maintain a statistically minimal risk of fuel exhaustion could result in yearly savings of 338 million lbs of CO 2 , the equivalent to the fuel consumed from 4760 flights on midsized commercial aircraft. Moreover, policy changes regarding maximum fuel loads or investments that reduce uncertainty or

  17. Environmental enrichment reduces innate anxiety with no effect on depression-like behaviour in mice lacking the serotonin transporter.

    Science.gov (United States)

    Rogers, Jake; Li, Shanshan; Lanfumey, Laurence; Hannan, Anthony J; Renoir, Thibault

    2017-08-14

    Along with being the main target of many antidepressant medications, the serotonin transporter (5-HTT) is known to be involved in the pathophysiology of depression and anxiety disorders. In line with this, mice with varying 5-HTT genotypes are invaluable tools to study depression- and anxiety-like behaviours as well as the mechanisms mediating potential therapeutics. There is clear evidence that both genetic and environmental factors play a role in the aetiology of psychiatric disorders. In that regard, housing paradigms which seek to enhance cognitive stimulation and physical activity have been shown to exert beneficial effects in animal models of neuropsychiatric disorders. In the present study, we examined the effects of environmental enrichment on affective-like behaviours and sensorimotor gating function of 5-HTT knock-out (KO) mice. Using the elevated-plus maze and the light-dark box, we found that environmental enrichment ameliorated the abnormal innate anxiety of 5-HTT KO mice on both tests. In contrast, environmental enrichment did not rescue the depression-like behaviour displayed by 5-HTT KO mice in the forced-swim test. Finally, measuring pre-pulse inhibition, we found no effect of genotype or treatment on sensorimotor gating. In conclusion, our data suggest that environmental enrichment specifically reduces innate anxiety of 5-HTT KO mice with no amelioration of the depression-like behaviour. This has implications for the current use of clinical interventions for patients with symptoms of both anxiety and depression. Copyright © 2017 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    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. (paper)

  19. Safe use of the Institute of Nuclear Physics reactor with low enriched fuel

    International Nuclear Information System (INIS)

    Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Salikhbaev, U.S.; Khalikov, U.A.; Yuldashev, B.S.

    2006-01-01

    Full text: The requirements for safe exploitation of reactor do not accept boiling of water on the surface of fuel elements. At determination of safe thermal regime of reactor (permissible level of power) the regime of the most heat-stressed fuel assembly (FA) in the active core was analyzed. By using ASTRA code [1] the heat-stressed sector is determined by most heat-stressed FA. In calculations the power of reactor was selected so that stock factor prior to the water boiling on the FA surface was not less than 1.45. Besides, in calculations the value of maximal energy density in examined FA is decreased by 10 %. As the part of the energy generated in the FA cores will be lost in constructional materials of the active zone and on the reflector. The stocks of safety before occurrence of instability of flow in gaps between of FA and before crisis of heat exchange are also analyzed. Further, by using the MCNP-4C code [2], densities of fast (E > 0,821 MeV) and thermal flows (E < 0,625 eV) of neutrons were calculated for those experimental channels where the irradiation of samples would be carried out. (author)

  20. 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)

    Montes T, J.L.; Cortes C, C.C.

    1992-08-01

    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)

  1. Microbial Diversity in Sulfate-Reducing Marine Sediment Enrichment Cultures Associated with Anaerobic Biotransformation of Coastal Stockpiled Phosphogypsum (Sfax, Tunisia

    Directory of Open Access Journals (Sweden)

    Hana Zouch

    2017-08-01

    Full Text Available Anaerobic biotechnology using sulfate-reducing bacteria (SRB is a promising alternative for reducing long-term stockpiling of phosphogypsum (PG, an acidic (pH ~3 by-product of the phosphate fertilizer industries containing high amounts of sulfate. The main objective of this study was to evaluate, for the first time, the diversity and ability of anaerobic marine microorganisms to convert sulfate from PG into sulfide, in order to look for marine SRB of biotechnological interest. A series of sulfate-reducing enrichment cultures were performed using different electron donors (i.e., acetate, formate, or lactate and sulfate sources (i.e., sodium sulfate or PG as electron acceptors. Significant sulfide production was observed from enrichment cultures inoculated with marine sediments, collected near the effluent discharge point of a Tunisian fertilizer industry (Sfax, Tunisia. Sulfate sources impacted sulfide production rates from marine sediments as well as the diversity of SRB species belonging to Deltaproteobacteria. When PG was used as sulfate source, Desulfovibrio species dominated microbial communities of marine sediments, while Desulfobacter species were mainly detected using sodium sulfate. Sulfide production was also affected depending on the electron donor used, with the highest production obtained using formate. In contrast, low sulfide production (acetate-containing cultures was associated with an increase in the population of Firmicutes. These results suggested that marine Desulfovibrio species, to be further isolated, are potential candidates for bioremediation of PG by immobilizing metals and metalloids thanks to sulfide production by these SRB.

  2. Reconfiguration of photovoltaic panels for reducing the hydrogen consumption in fuel cells of hybrid systems

    Directory of Open Access Journals (Sweden)

    Daniel González-Montoya

    2017-05-01

    Full Text Available Hybrid generation combines advantages from fuel cell systems with non-predictable generation approaches, such as photovoltaic and wind generators. In such hybrid systems, it is desirable to minimize as much as possible the fuel consumption, for the sake of reducing costs and increasing the system autonomy. This paper proposes an optimization algorithm, referred to as population-based incremental learning, in order to maximize the produced power of a photovoltaic generator. This maximization reduces the fuel consumption in the hybrid aggregation. Moreover, the algorithm's speed enables the real-time computation of the best configuration for the photovoltaic system, which also optimizes the fuel consumption in the complementary fuel cell system. Finally, a system experimental validation is presented considering 6 photovoltaic modules and a NEXA 1.2KW fuel cell. Such a validation demonstrates the effectiveness of the proposed algorithm to reduce the hydrogen consumption in these hybrid systems.

  3. Healthy reduced-fat Bologna sausages enriched in ALA and DHA and stabilized with Melissa officinalis extract.

    Science.gov (United States)

    Berasategi, Izaskun; Navarro-Blasco, Iñigo; Calvo, Maria Isabel; Cavero, Rita Yolanda; Astiasarán, Iciar; Ansorena, Diana

    2014-03-01

    Reduced-energy and reduced-fat Bologna products enriched with ω-3 polyunsaturated fatty acids were formulated by replacing the pork back-fat by an oil-in-water emulsion containing a mixture of linseed-algae oil stabilized with a lyophilized Melissa officinalis extract. Healthier composition and lipid profile was obtained: 85 kcal/100 g, 3.6% fat, 0.6 g ALA and 0.44 g DHA per 100 g of product and ω-6/ω-3 ratio of 0.4. Technological and sensory problems were not detected in the new formulations. Reformulation did not cause oxidation problems during 32 days of storage under refrigeration. The results suggest that it is possible to obtain reduced-fat Bologna-type sausages rich in ALA and DHA and stabilized with natural antioxidants, applying the appropriate technology without significant effects on the sensory quality, yielding interesting products from a nutritional point of view. © 2013.

  4. Hydrogen as a renewable and sustainable solution in reducing global fossil fuel consumption

    International Nuclear Information System (INIS)

    Midilli, Adnan; Dincer, Ibrahim

    2008-01-01

    In this paper, hydrogen is considered as a renewable and sustainable solution for reducing global fossil fuel consumption and combating global warming and studied exergetically through a parametric performance analysis. The environmental impact results are then compared with the ones obtained for fossil fuels. In this regard, some exergetic expressions are derived depending primarily upon the exergetic utilization ratios of fossil fuels and hydrogen: the fossil fuel based global waste exergy factor, hydrogen based global exergetic efficiency, fossil fuel based global irreversibility coefficient and hydrogen based global exergetic indicator. These relations incorporate predicted exergetic utilization ratios for hydrogen energy from non-fossil fuel resources such as water, etc., and are used to investigate whether or not exergetic utilization of hydrogen can significantly reduce the fossil fuel based global irreversibility coefficient (ranging from 1 to +∞) indicating the fossil fuel consumption and contribute to increase the hydrogen based global exergetic indicator (ranging from 0 to 1) indicating the hydrogen utilization at a certain ratio of fossil fuel utilization. In order to verify all these exergetic expressions, the actual fossil fuel consumption and production data are taken from the literature. Due to the unavailability of appropriate hydrogen data for analysis, it is assumed that the utilization ratios of hydrogen are ranged between 0 and 1. For the verification of these parameters, the variations of fossil fuel based global irreversibility coefficient and hydrogen based global exergetic indicator as the functions of fossil fuel based global waste exergy factor, hydrogen based global exergetic efficiency and exergetic utilization of hydrogen from non-fossil fuels are analyzed and discussed in detail. Consequently, if exergetic utilization ratio of hydrogen from non-fossil fuel sources at a certain exergetic utilization ratio of fossil fuels increases

  5. Shielding Studies for Reducing the associated Radiological Risks Due To Irradiated Low Enriched Uranium Foil

    International Nuclear Information System (INIS)

    Margeanu, C.A.

    2011-01-01

    Present work estimates the radiation dose rates corresponding to irradiated Low Enriched Uranium (20 wt % 235 U) foil as part of shielding studies for radiological risks reduction after irradiation inside TRIGA 14 MW Research Reactor in an investigation on 99 Mo production possibility. Post-Irradiation Examination Laboratory's cell shielding calculations have been performed; radiation source was obtained by using ORIGEN-S code with specific cross-sections libraries. Different post-irradiation cooling times have been considered, gamma dose rates being estimated by using MAVRIC module from Scale 6 programs package, for following exposure situations (relative to Pie cell): i) front side, ii) lateral side and iii) back side. Three different calculations were performed: a) without any protection shield between operator and cell, except for the cell stainless steel wall; b) with a Lead protection shield between operator and cell and c) with a depleted Uranium shield, located inside the cell in between the radiation source and cell window. Radiation dose rates to cell external wall surface and for other eight fixed distances from cell wall were estimated. To obtain a consistent set of solutions, the study was done for various Uranium foil weights and different Lead and depleted Uranium shields thicknesses. Calculations were focused to assure that the dose rate to an operator positioned at 60 cm working distance from the cell will not exceed 0.02 mSv/h, maximum allowed dose rate for professionally exposed personnel according to Romanian regulations.

  6. 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)

  7. 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)

    1998-12-31

    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)

  8. On-line item control at a high enriched nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Lewis, T.W.; Lewis, H.M.

    1984-01-01

    The on-line item control system at Nuclear Fuel Services, Inc., is a near-real time method capable of tracking uniquely identified items from creation through disposition. The system provides for improved control, timeliness, accuracy and usability of company information and the necessary data required to support the regulatory program for the protection against diversion of Special Nuclear Materials. The system consists of software applications (approximately 150 programs) with man/machine interface controls which provide facilities for correct data entry and for the protection of data integrity. This system went into stand-alone operation in September, 1983 after a twenty month parallel test run with the previous keybatched (manual forms) item control system

  9. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  10. The effectiveness of enriching relations between spouses to reduce marital conflict between employees in different offices in Yasouj

    Directory of Open Access Journals (Sweden)

    M Aminianfar

    2015-09-01

    Full Text Available Objective & aim:  Today, divorce and marital turmoil is increasing. Understanding the factors leading to chaos and the disintegration of family relationships is important. This study aimed to explore the effect of enriching relations between spouses, the couple's emotional security and marital conflicts.   Methods: In the present interventional-analytical study, ten different offices were randomly selected.  Of people who scored high on marital conflict and emotional security, and also those who gained low scores under 40 were divided randomly into two groups. Seven relations enriching group training sessions were held for the spouses.  At the end of the training sessions, both groups were evaluated by emotional security questionnaires by Brunner et al. (2008, marital conflict Sanaei and Barati (1996.   Results: Covariance analysis and multivariate analysis of variance, analysis of the results of the marital conflict, mean and standard deviation of pre couples' marital conflict experimental and control groups were (20/15 and 20/145 (64/16 70/143 respectively.  The test score in two groups were (60/12 and 80/64 (17.4 and 70/143 respectively. The results of multivariate analysis of covariance components of marital conflict on test scores of experimental and control groups and the control effect of pre-test showed that Pylayy effect, Wilks Lambda test, Hotelling effect on the root of F=4.47 and degrees of freedom 7 levels significantly in p=0.0001 Effect of married couples was significant in reducing aggression.   Conclusion: Enrichment relations education for spouses may significantly reduce parameters of marital conflicts.

  11. 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)

  12. From high to low. The IAEA is helping to reduce the use of high-risk nuclear fuel at the world's research reactors

    International Nuclear Information System (INIS)

    Adelfang, P.; Goldman, I.

    2006-01-01

    Research reactors play a key role in the development of peaceful uses of atomic energy. They are used for the production of isotopes for medicine and industry, for research in physics, biology and materials science, and for scientific education and training. They also continue to play an important role in support of nuclear power programmes. The IAEA's data shows there are 249 operational research reactors worldwide. Of these, more than 100 reactors are still fuelled with highly enriched uranium (HEU). It is considered high-risk nuclear material since it can be easily used for a nuclear explosive device. As part of a developing international norm to minimize and eventually eliminate HEU in civilian nuclear applications, research reactor operators increasingly are working with national and international agencies. They are being encouraged and supported to improve their physical security arrangements, convert their reactors to low-enriched uranium (LEU) fuel, and ship irradiated fuel back to the country of origin.For more than twenty years the IAEA has been supporting international efforts associated with reducing the amount of HEU in international commerce. Projects and activities have directly supported a programme the United States initiated in 1978, called Reduced Enrichment for Research and Test Reactors (RERTR). The IAEA's work additionally supports efforts to return research reactor fuel to the country where it was originally enriched so-called take back activities. IAEA initiatives have included the development and maintenance of several databases with information related to research reactors and research reactor spent fuel inventories. These databases have been essential in planning and managing both RERTR and take-back programmes. Other Agency activities through technical cooperation and other channels have supported the conversion of research reactors to using lower enriched fuels. In other ways, the IAEA supports the exchange of information among experts

  13. N2O-reducing activity of soil amended with organic and inorganic enrichments under flooded conditions

    Directory of Open Access Journals (Sweden)

    Alicja Księżopolska

    Full Text Available ABSTRACT Changes, apparent after investigation, in the physical and chemical properties in soil, as a result of organic and inorganic enrichments under flooded conditions, influence the growth of denitrifiers. The aim of this study was to determine the effect of the addition of manure (8 kg m−2 (M, clay (50 kg m−2 (CL and lime (1.12 kg m−2 (Ca on the N2O-reducing activity (N2O-RA of sandy loam soil (clay content - 24 % in 0-20 cm, during NO3 reduction under flooding. The soil samples were taken from field plots after 3 years of enrichment with grass cultivation. The enrichments had a distinct effect on the N2O-RA and N2O-released, due to the change in pH, the porosity, and the sorptive properties of the soil. The pH had the greatest impact on the N2O-RA of the soil and ranged from 4.9 to 7.6. For actual denitrification to N2O-realized (aD-N2O, the maximum N2O-releasing (mcN2O-releasing followed the order: 1.36 for the M-treatment, 6.39 for the M+CL+Ca-treatment, 7.79 for the c-soil and 8.69 N2O-N mg kg−1 for the M+CL-treatment. For actual denitrification (aD, the mcN2O-releasing was followed the order: 10.37 for the M-treatment, 10.49 for the control soil, 14.60 for the M+CL+Ca-treatment and 20.00 N2O-N mg kg−1 for the M+CL-treatment. The N2O-RA of the soil samples increased as pH increased. The average N2O/N2+N2O ratio and the N2O-RA of the soil samples increased in the following order: M+CL, control soil, M+CL+Ca, M-enrichments. The addition of enrichments did not pose a threat to the environment due to increased N2O emissions, but as regards conserving NO3− in the soil, the addition of clay distinctly increased the complete denitrification process.

  14. Direct methanol feed fuel cell with reduced catalyst loading

    Science.gov (United States)

    Kindler, Andrew (Inventor)

    1999-01-01

    Improvements to direct feed methanol fuel cells include new protocols for component formation. Catalyst-water repellent material is applied in formation of electrodes and sintered before application of ionomer. A membrane used in formation of an electrode assembly is specially pre-treated to improve bonding between catalyst and membrane. The improved electrode and the pre-treated membrane are assembled into a membrane electrode assembly.

  15. Multifunctional Fuel Additives for Reduced Jet Particulate Emissions

    Science.gov (United States)

    2006-06-01

    Propulsion, Santiago , Chile , Mar. 8-11, 2005. Montgomery, C. J., Sarofim, A. F., Preciado, I., Marsh, N. D., Eddings, E. G., and Bozzelli, J. W...34Temperature and CO2 concentration measurements in the exhaust stream of a liquid- fueled combustor using dual-pump coherent anti-Stokes Raman scattering...injection pressure, and oxygen concentration . Additives were found to be most effective under highly oxidizing conditions. Soot reductions of over 90% were

  16. Optimization of neutronic characteristics of U3Si2 low enrichment fuel elements for a new design of IEA-R1 reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.; Maiorino, J.R.; Gouvea, E.A.

    1989-01-01

    This work shows a study of neutronic optimization of U 3 Si 2 -Al low enrichment fuel element. This study has a goal to propose a optimized Core to be used in the research reactor IEA-R1. The external dimensions of the fuel element were maintained as constraints and the loss of reactivity along fuel life-time was defined as 'objective function', and it has been minimized by varying the fuel element dimensions. Cell calculations were made with HAMMER-TECH /3/ Code, for burnups up to 50% of U-235 initial mass. The Computer values of the objective function for several combinations of fuel element dimensions were fitted by a surface using the SAS system /9/, and it has been minimized by a Harwell subroutine /10/. (author) [pt

  17. 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)

    Pechin, W.H.; Blanco, R.E.; Dahlman, R.C.; Finney, B.C.; Lindauer, R.B.; Witherspoon, J.P.

    1975-05-01

    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.)

  18. HEATHYD, Steady-State Thermal Hydraulic Analysis of Low-Enriched U Fuel Reactor

    International Nuclear Information System (INIS)

    NABBI, R.

    1989-01-01

    1 - Description of program or function: HEATHYD is a code for the steady-state heat transfer calculation of research nuclear reactors with forced convection. It models heat transfer and coolant flow for assemblies of parallel fuel plates of MTR type with any axial power distribution. The thermodynamic model accounts for single phase cooling and sub- cooled boiling condition using the transition criterion of Bergeles-Rosenow. In addition to the calculation of the channel flow velocities and coolant pressure drops, HEATHYD calculates axial distribution of the coolant and clad-surface temperatures. Safety margins to the critical heat flux as a result of burnout condition or flow instability are determined. 2 - Method of solution: Applying the finite difference method, HEATHYD solves the equations of heat conduction and heat transfer to the coolant. For the physical properties of the coolant as a function of the coolant temperature polynomials of degree 6 are used. Depending on the coolant condition, different correlations for the heat transfer coefficient can be applied. The analysis of the critical cooling conditions resulting in burnout or flow instability, is performed according to the correlations developed by Mirshak/ Labuntsov and Forgan/Whittle

  19. Forest fuel reduces the nitrogen load - calculations of nitrogen flows

    International Nuclear Information System (INIS)

    Burstroem, F.; Johansson, Jan.

    1995-12-01

    Nitrogen deposition in Sweden has increased strongly during recent decades, particularly in southern Sweden. Nitrogen appears to be largely accumulated in biomass and in the soil. It is therefore desirable to check the accumulation of nitrogen in the forest. The most suitable way of doing this is to remove more nitrogen-rich biomass from the forest, i.e., increase the removal of felling residues from final fellings and cleanings. An ecological condition for intensive removal of fuel is that the ashes are returned. The critical load for nitrogen, CL(N), indicates the level of nitrogen deposition that the forest can withstand without leading to ecological changes. Today, nitrogen deposition is higher than the CL(N) in almost all of Sweden. CL(N) is calculated in such a manner that nitrogen deposition should largely be balanced by nitrogen losses through harvesting during a forest rotation. The value of CL(N) thus largely depends on how much nitrogen is removed with the harvested biomass. When both stems and felling residues are harvested, the CL(N) is about three times higher than in conventional forestry. The increase is directly related to the amount of nitrogen in the removed biofuel. Use of biofuel also causes a certain amount of nitrogen emissions. From the environmental viewpoint there is no difference between the sources of the nitrogen compounds. An analysis of the entire fuel chain shows that, compared with the amount of nitrogen removed from the forest with the fuel, about 5 % will be emitted as nitrogen oxides or ammonia during combustion, and a further ca 5 % during handling and transports. A net amount of about 90 % of biomass nitrogen is removed from the system and becomes inert nitrogen (N 2 ). 60 refs, 3 figs, 4 tabs, 11 appendices

  20. Solid oxide fuel cell cathode with oxygen-reducing layer

    Science.gov (United States)

    Surdoval, Wayne A.; Berry, David A.; Shultz, Travis

    2018-04-03

    The disclosure provides a SOFC comprised of an electrolyte, anode, and cathode, where the cathode comprises an MIEC and an oxygen-reducing layer. The oxygen-reducing layer is in contact with the MIEC, and the MIEC is generally between and separating the oxygen-reducing layer and the electrolyte. The oxygen-reducing layer is comprised of single element oxides, single element carbonates, or mixtures thereof, and has a thickness of less than about 30 nm. In a particular embodiment, the thickness is less than 5 nm. In another embodiment, the thickness is about 3 monolayers or less. The oxygen-reducing layer may be a continuous film or a discontinuous film with various coverage ratios. The oxygen-reducing layer at the thicknesses described may be generated on the MIEC surface using means known in the art such as, for example, ALD processes.

  1. Renewable and nuclear sources of energy reduce the share of fossil fuels

    International Nuclear Information System (INIS)

    Koprda, V.

    2009-01-01

    In this paper author presents a statistical data use of nuclear energy, renewable sources and fossil fuels in the share of energy production in the Slovak Republic. It is stated that use of nuclear energy and renewable sources reduce the share of fossil fuels.

  2. Evaluating alternative fuel treatment strategies to reduce wildfire losses in a Mediterranean area

    Science.gov (United States)

    Michele Salis; Maurizio Laconi; Alan A. Ager; Fermin J. Alcasena; Bachisio Arca; Olga Lozano; Ana Fernandes de Oliveira; Donatella Spano

    2016-01-01

    The goal of this work is to evaluate by a modeling approach the effectiveness of alternative fuel treatment strategies to reduce potential losses from wildfires in Mediterranean areas. We compared strategic fuel treatments located near specific human values vs random locations, and treated 3, 9 and 15% of a 68,000 ha study area located in Sardinia, Italy. The...

  3. Fuel treatment effectiveness in reducing fire intensity and spread rate - An experimental overview

    Science.gov (United States)

    Eric Mueller; Nicholas Skowronski; Albert Simeoni; Kenneth Clark; Robert Kremens; William Mell; Michael Gallagher; Jan Thomas; Alexander Filkov; Mohamad El Houssami; John Hom; Bret Butler

    2014-01-01

    Fuel treatments represent a significant component of the wildfire mitigation strategy in the United States. However, the lack of research aimed at quantifying the explicit effectiveness of fuel treatments in reducing wildfire intensity and spread rate limits our ability to make educated decisions about the type and placement of these treatments. As part of a larger...

  4. Uranium Enrichment, an overview

    International Nuclear Information System (INIS)

    Coates, J.H.

    1994-01-01

    This general presentation on uranium enrichment will be followed by lectures on more specific topics including descriptions of enrichment processes and assessments of the prevailing commercial and industrial situations. I shall therefore avoid as much as possible duplications with these other lectures, and rather dwell on: some theoretical aspects of enrichment in general, underlying the differences between statistical and selective processes, a review and comparison between enrichment processes, remarks of general order regarding applications, the proliferation potential of enrichment. It is noteworthy that enrichment: may occur twice in the LWR fuel cycle: first by enriching natural uranium, second by reenriching uranium recovered from reprocessing, must meet LWR requirements, and in particular higher assays required by high burn up fuel elements, bears on the structure of the entire front part of the fuel cycle, namely in the conversion/reconversion steps only involving UF 6 for the moment. (author). tabs., figs., 4 refs

  5. Development of ISA procedure for uranium fuel fabrication and enrichment facilities: overview of ISA procedure and its application

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki

    2013-01-01

    Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)

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

    International Nuclear Information System (INIS)

    Burbidge, B.L.H.; Franklin, B.M.; Small, V.G.

    1983-01-01

    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)

  7. In situ electrochemical enrichment and isolation of a magnetite-reducing bacterium from a high pH serpentinizing spring.

    Science.gov (United States)

    Rowe, Annette R; Yoshimura, Miho; LaRowe, Doug E; Bird, Lina J; Amend, Jan P; Hashimoto, Kazuhito; Nealson, Kenneth H; Okamoto, Akihiro

    2017-06-01

    Serpentinization is a geologic process that produces highly reduced, hydrogen-rich fluids that support microbial communities under high pH conditions. We investigated the activity of microbes capable of extracellular electron transfer in a terrestrial serpentinizing system known as 'The Cedars'. Measuring current generation with an on-site two-electrode system, we observed daily oscillations in current with the current maxima and minima occurring during daylight hours. Distinct members of the microbial community were enriched. Current generation in lab-scale electrochemical reactors did not oscillate, but was correlated with carbohydrate amendment in Cedars-specific minimal media. Gammaproteobacteria and Firmicutes were consistently enriched from lab electrochemical systems on δ-MnO 2 and amorphous Fe(OH) 3 at pH 11. However, isolation of an electrogenic strain proved difficult as transfer cultures failed to grow after multiple rounds of media transfer. Lowering the bulk pH in the media allowed us to isolate a Firmicutes strain (Paenibacillus sp.). This strain was capable of electrode and mineral reduction (including magnetite) at pH 9. This report provides evidence of the in situ activity of microbes using extracellular substrates as sinks for electrons at The Cedars, but also highlights the potential importance of community dynamics for supporting microbial life through either carbon fixation, and/or moderating pH stress. © 2017 The Authors. Environmental Microbiology published by Society for Applied Microbiology and John Wiley & Sons Ltd.

  8. Reduced fuel consumption for fork-lift trucks with hydrostatic transmission

    Energy Technology Data Exchange (ETDEWEB)

    Abels, T

    1983-05-01

    Cost calculations for a 3,5-t diesel fork lifter done on the basis of VDI 2695 shows, that fuel costs account only for a small part of the operating costs despite the price increase for diesel fuel. Fork lifters with disk-cam controlled primary/secondary adjusted hydrostatic transmission used less fuel than was indicated in the VDI-guideline. Fuel consumption could further be reduced by an optimized hydraulic adjustment together with a precisely harmonized engine speed adjustment. Annual cost savings are considerable.

  9. Evaluation of Techniques for Reducing In-Use Automotive Fuel Consumption

    Science.gov (United States)

    1981-04-01

    This report presents an assessment of proposed techniques for reducing fuel consumption in the in-use light duty road vehicle fleet. Three general classes of techniques are treated: (1) modification of vehicles, (2) modification of traffic flow, and ...

  10. Maillard reaction products enriched food extract reduce the expression of myofibroblast phenotype markers.

    Science.gov (United States)

    Ruhs, Stefanie; Nass, Norbert; Somoza, Veronika; Friess, Ulrich; Schinzel, Reinhard; Silber, Rolf-Edgar; Simm, Andreas

    2007-04-01

    Advanced glycation end products (AGE) are associated with a wide range of degenerative diseases. The present investigation aimed at analysing the influence of AGE containing nutritional extracts on cardiac fibroblasts (CFs) as the major cell type responsible for cardiac fibrosis. Mice CFs were treated with bread crust extract (BCE) which contained significant amounts of a variety of AGE modifications. BCE treatment with up to 30 mg/mL did not impair cell viability. Furthermore, BCE induced a moderate elevation of reactive oxygen species (ROS) production and activation of redox sensitive pathways like the p42/44(MAPK), p38(MAPK) and NF-kappaB but did not alter Akt kinase phosphorylation. Expression of smooth muscle alpha-actin and tropomyosin-1, which represent markers for myofibroblast differentiation, was reduced after bread crust treatment. These data suggest a putative antifibrotic effect of melanoidin-rich food.

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

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

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

  12. Experimental RA reactor operation with 80% enriched fuel - Program of experimental operation: a) Program of experimental operation with 80% enriched fuel at low power, b) contents of the experimental operation with 80% enriched fuel at higher power levels; Program probnog rada: a) Program probnog rada reaktora sa 80% obogacenim gorivom na malim snagama, b) sadrzaj programa probnog rada reaktora RA sa 80% obogacenim gorivom na vecim snagama

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Sotic, O; Skoric, M; Cupac, S; Bulovic, V; Maric, I; Marinkov, L [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-10-15

    Highly enriched (80%) uranium oxide fuel was regularly used in the mixed reactor core with the 2% enriched fuel since 1976. The most important changes related to reactor operation, in comparison with the original design project were related to reactor core fuelling schemes. At the end of 1979 reactor was shutdown due to the corrosion coating noticed on some fuel elements and due to decrease quality of the heavy water. Subsequently the Sanitary inspector of Serbia has prohibited further reactor operation. Restart of the reactor will not be a simple continuation of operation. It is indispensable to perform complete experimental program including measurements of critical parameters at different power levels for the core with fresh 80% enriched fuel. The aim of this document is to obtain working permission and its contents are in agreement with the procedure demanded by the Safety Committee of the Institute. It includes results of optimization and safety analysis for the initial reactor core. Since the permission for restart is not obtained, a separate RA reactor safety report is prepared in addition to the program for experimental operation. This report includes: detailed program for reactor experimental operation with 80% enriched fuel in the core at low power levels, and contents of the experimental operation with 80% enriched fuel in the core at higher power levels. [Serbo-Croat] Od decembra 1976. godine redovno je korisceno 80% obogaceno gorivo u mesanoj resetki reaktorskog jezgra sa 2% obogacenim gorivom. Najvece izmene na reaktoru u odnosu na originalni projekat izvrsene su u nacinu rukovanja gorivom. Krajem marta 1979. godine obustavljen je rad reaktora usled naslaga na gorivnim elementima i loseg stanja teske vode. Naknadno je izdata zabrana za rad reaktora od strane Sanitarnog inspektora SR Srbije. Ponovno pustanje reaktora u rad nece biti jednostavan nastavak rada. Neophodno je da se izvede kompletan program merenja kriticnih parametara i drugih

  13. A study on items necessary to develop the requirements for the management of serious accidents postulated in fuel fabrication, enrichment and reprocessing facilities

    International Nuclear Information System (INIS)

    Takanashi, Mitsuhiro; Yamate, Kazuki; Asada, Kazuo; Yamada, Takashi; Endo, Shigeki

    2013-05-01

    The purpose of this study is to supply the points to discuss on new rules of fuel fabrication, enrichment and reprocessing facilities (hereinafter referred to as 'fuel cycle facilities') conducted by Nuclear Regulation Authority. Requirements for management of serious accidents in the fuel cycle facilities were summarized in this study. Taking into account the lessons learned from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant in Mar. 2011, Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors was amended in June 2012. The main items of the amendment were as follows: Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facilities (back-fitting). Japan Nuclear Energy Safety organization (JNES) conducted a fundamental study on serious accidents and their management in the fuel cycle facilities and made a report. In the report, the concept of Defense in Depth and the definition of serious accidents for the fuel cycle facilities were discussed. Those discussions were conducted by reference to new regulation rules (draft) for power reactors and from the view of features of the fuel cycle facilities. However, further detailed studies are necessary in order to clarify some issues in it. It was also reflected opinions from experts in JNES technical meetings on accident management of the fuel cycle facilities to brush up this report. (author)

  14. Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

    2015-01-01

    Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

  15. 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.

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

    International Nuclear Information System (INIS)

    Primm, R.T. III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-01-01

    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 3 O 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

  17. REDUCING ULTRA-CLEAN TRANSPORTATION FUEL COSTS WITH HYMELT HYDROGEN

    Energy Technology Data Exchange (ETDEWEB)

    Donald P. Malone; William R. Renner

    2003-07-31

    This report describes activities for the third quarter of work performed under this agreement. Atmospheric testing was conducted as scheduled on June 5 through June 13, 2003. The test results were encouraging, however, the rate of carbon dissolution was below expectations. Additional atmospheric testing is scheduled for the first week of September 2003. Phase I of the work to be done under this agreement consists of conducting atmospheric gasification of coal using the HyMelt technology to produce separate hydrogen rich and carbon monoxide rich product stream. In addition smaller quantities of petroleum coke and a low value refinery stream will be gasified. DOE and EnviRes will evaluate the results of this work to determine the feasibility and desirability of proceeding to Phase II of the work to be done under this agreement, which is gasification of the above-mentioned feeds at a gasifier pressure of approximately 5 bar. The results of this work will be used to evaluate the technical and economic aspects of producing ultra-clean transportation fuels using the HyMelt technology in existing and proposed refinery configurations.

  18. Reducing the fuel use and greenhouse gas emissions of the US vehicle fleet

    International Nuclear Information System (INIS)

    Bandivadekar, Anup; Cheah, Lynette; Evans, Christopher; Groode, Tiffany; Heywood, John; Kasseris, Emmanuel; Kromer, Matthew; Weiss, Malcolm

    2008-01-01

    The unrelenting increase in the consumption of oil in the US light-duty vehicle fleet (cars and light trucks) presents an extremely challenging energy and environmental problem. A variety of propulsion technologies and fuels have the promise to reduce petroleum use and greenhouse gas emissions from motor vehicles. Even so, achieving a noticeable reduction on both fronts in the near term will require rapid penetration of these technologies into the vehicle fleet, and not all alternatives can meet both objectives simultaneously. Placing a much greater emphasis on reducing fuel consumption rather than improving vehicle performance can greatly reduce the required market penetration rates. Addressing the vehicle performance-size-fuel consumption trade-off should be the priority for policymakers rather than promoting specific vehicle technologies and fuels

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

    International Nuclear Information System (INIS)

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-01-01

    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) (σ 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).

  20. Neutronic calculations with transport and diffusion computer codes for light water moderated critical with UO2 enriched at 4,75% as fuel

    International Nuclear Information System (INIS)

    Sabundjian, G.; Nakata, H.

    1983-02-01

    The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt

  1. Future of uranium enrichment

    International Nuclear Information System (INIS)

    Hosmer, C.

    1981-01-01

    The increasing amount of separative work being done in government facilities to produce low-enriched uranium fuel for nuclear utilities again raises the question: should this business-type, industrial function be burned over the private industry. The idea is being looked at by the Reagan administration, but faces problems of national security as well as from the unique nature of the business. This article suggests that a joint government-private venture combining enriching, reprocessing, and waste disposal could be the answer. Further, a separate entity using advanced laser technology to deplete existing uranium tails and lease them for fertile blankets in breeder reactors might earn substantial revenues to help reduce the national debt

  2. Economic feasibility of hydrogen enrichment for reducing NOx emissions from landfill gas power generation alternatives: A comparison of the levelized cost of electricity with present strategies

    International Nuclear Information System (INIS)

    Kornbluth, Kurt; Greenwood, Jason; Jordan, Eddie; McCaffrey, Zach; Erickson, Paul A.

    2012-01-01

    Based on recent research showing that hydrogen enrichment can lower NO x emissions from landfill gas combustion below future NO x emission control standards imposed by both federal and California state regulations, an investigation was performed to compare the levelized cost of electricity of this technology with other options. In this cost study, a lean-burn reciprocating engine with no after-treatment was the baseline case to compare six other landfill gas-to-energy projects. These cases include a lean burn engine with selective catalytic reduction after treatment, a lean-burn microturbine, and four variations on an ultra-lean-burn engine utilizing hydrogen enrichment with each case using a different method of hydrogen production. Only hydrogen enrichment with an in-stream autothermal fuel reformer was shown to be potentially cost-competitive with current strategies for reaching the NO x reduction target in IC engines. - Highlights: ► Levelized cost of electricity for hydrogen enriched combustion was compared. ► Various ultra-lean-burn engines and microturbines with hydrogen were analyzed. ► Combustion with an autothermal fuel reformer was potentially cost-competitive.

  3. Reducing fuel subsidies and the implication on fiscal balance and poverty in Indonesia: A simulation analysis

    International Nuclear Information System (INIS)

    Dartanto, Teguh

    2013-01-01

    There is an urgent need for phasing out fuel subsidies in Indonesia due to a severe budget deficit and a worsening of income distribution. Fuel subsidies, of which almost 72% are enjoyed by the 30% of the richest income groups, have consumed on average 63.8% of the total subsidies between 1998 and 2013. This paper aims at evaluating the relationship between existing fuel subsidies and fiscal balance and at analysing the poverty impact of phasing out fuel subsidies. Applying a CGE-microsimulation, this study found that removing 25% of fuel subsidies increases the incidence of poverty by 0.259 percentage points. If this money were fully allocated to government spending, the poverty incidence would decrease by 0.27 percentage points. Moreover, the 100% removal of fuel subsidies and the reallocation of 50% of them to government spending, transfers and other subsidies could decrease the incidence of poverty by 0.277 percentage points. However, these reallocation policies might not be effective in compensating for the adverse impacts of the 100% removal of fuel subsidies if economic agents try to seek profit through mark-up pricing over the increase of production costs. - Highlights: ► Massive fuel subsidies reduce fiscal spaces used to alleviate poverty in Indonesia. ► Indonesia can avoid a budget deficit by 78% cutting of fuel subsidies. ► A CGE-microsimulation is applied to analyse the impacts of fuel subsidy reallocation. ► The 50% of reallocation fuel subsidies decreases the poverty by 0.277 percentage points. ► Mark-up pricing done by economic agents reduces the effectiveness of reallocation

  4. Portulaca oleracea reduces triglyceridemia, cholesterolemia, and improves lecithin: cholesterol acyltransferase activity in rats fed enriched-cholesterol diet.

    Science.gov (United States)

    Zidan, Y; Bouderbala, S; Djellouli, F; Lacaille-Dubois, M A; Bouchenak, M

    2014-10-15

    The effects of Portulaca oleracea (Po) lyophilized aqueous extract were determined on the serum high-density lipoproteins (HDL2 and HDL3) amounts and composition, as well as on lecithin: cholesterol acyltansferase (LCAT) activity. Male Wistar rats (n = 12) were fed on 1% cholesterol-enriched diet for 10 days. After this phase, hypercholesterolemic rats (HC) were divided into two groups fed the same diet supplemented or not with Portulaca oleracea (Po-HC) (0.5%) for four weeks. Serum total cholesterol (TC) and triacylglycerols (TG), and liver TG values were respectively 1.6-, 1.8-, and 1.6-fold lower in Po-HC than in HC group. Cholesterol concentrations in LDL-HDL1, HDL2, and HDL3 were respectively 1.8, 1.4-, and 2.4-fold decreased in Po-HC group. HDL2 and HDL3 amounts, which were the sum of apolipoproteins (apos), TG, cholesteryl esters (CE), unesterified cholesterol (UC), and phospholipids (PL) contents, were respectively 4.5-fold higher and 1.2-fold lower with Po treatment. Indeed, enhanced LCAT activity (1.2-fold), its cofactor-activator apo A-I (2-fold) and its reaction product HDL2-CE (2.1-fold) were observed, whereas HDL3-PL (enzyme substrate) and HDL3-UC (acyl group acceptor) were 1.2- and 2.4-fold lower. Portulaca oleracea reduces triglyceridemia, cholesterolemia, and improves reverse cholesterol transport in rat fed enriched-cholesterol diet, contributing to anti-atherogenic effects. Copyright © 2014 Elsevier GmbH. All rights reserved.

  5. Fructose-enriched diet induces inflammation and reduces antioxidative defense in visceral adipose tissue of young female rats.

    Science.gov (United States)

    Kovačević, Sanja; Nestorov, Jelena; Matić, Gordana; Elaković, Ivana

    2017-02-01

    The consumption of refined, fructose-enriched food continuously increases and has been linked to development of obesity, especially in young population. Low-grade inflammation and increased oxidative stress have been implicated in the pathogenesis of obesity-related disorders including type 2 diabetes. In this study, we examined alterations in inflammation and antioxidative defense system in the visceral adipose tissue (VAT) of fructose-fed young female rats, and related them to changes in adiposity and insulin sensitivity. We examined the effects of 9-week fructose-enriched diet applied immediately after weaning on nuclear factor κB (NF-κB) intracellular distribution, and on the expression of pro-inflammatory cytokines (IL-1β and TNFα) and key antioxidative enzymes in the VAT of female rats. Insulin signaling in the VAT was evaluated at the level of insulin receptor substrate-1 (IRS-1) protein and its inhibitory phosphorylation on Ser 307 . Fructose-fed rats had increased VAT mass along with increased NF-κB nuclear accumulation and elevated IL-1β, but not TNFα expression. The protein levels of antioxidative defense enzymes, mitochondrial manganese superoxide dismutase 2, and glutathione peroxidase, were reduced, while the protein content of IRS-1 and its inhibitory phosphorylation were not altered by fructose diet. The results suggest that fructose overconsumption-related alterations in pro-inflammatory markers and antioxidative capacity in the VAT of young female rats can be implicated in the development of adiposity, but do not affect inhibitory phosphorylation of IRS-1.

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  7. The Caramel fuel in OSIRIS

    International Nuclear Information System (INIS)

    Cherruau, Francois.

    1980-11-01

    This paper presents the main characteristics of the caramel fuel, a description of OSIRIS transformations that were decided in line with its conversion and the results of its operation since then. The Caramel fuel is made from sintered UO 2 pellets contained in zircaloy clads forming the plates of the fuel assembly reducing the enrichment need to as little as 3 to 10% instead of 93% enriched U/Al in the previous fuel. The first year of experience shows the capacity under a statistic scale of the caramel fuel to fulfil the most severe operation requirements for use in low and medium power research reactors

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

    1998-01-01

    In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

  9. Application of gamma spectrometry technique in combination with weighing for material balance taking in the production of highly enriched U-A1 fuel

    International Nuclear Information System (INIS)

    Serin, P.A.

    1975-07-01

    The purpose of this project is to obtain the data on material balance for a batch of highly enriched U-Al alloys (used in the NRX and NRU reactors) during production of fuel, using gamma spectrometry (mainly the 186 KeV photopeak) and weighing, and to determine operational data of the Agency's single channel stabilized spectrometer (SAM-1) for measurement of the product typical for the production of highly enriched U-Al fuel (U-Al billets, fuel elements, scrap). The data collected indicates that gamma spectrometry using the single channel stabilized spectrometer is a valid non-destructive method of determining quantitatively U-235 content of U-Al alloy in the form of cast billets or extruded fuel elements providing that adequate standards are available. An accuracy of better than + 1% relative can be obtained using a simple jig to provide reproducible counting geometry. Count rates should be kept well below the saturation level of the detector and counter, preferably by a lead collimator in front of the detector. This non-destructive method is not easily applicable to scrap because of the inability to maintain constant geometry and to prepare standards closely similar in size and shape to the samples

  10. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  11. Characterization of nitrate-reducing and amino acid-using bacteria prominent in nitrotoxin-enriched equine cecal populations

    Science.gov (United States)

    In the present study, populations of equine cecal microbes enriched for enhanced rates of 3-nitro-1-propionic acid (NPA) or nitrate metabolism were diluted and cultured for NPA-metabolizing bacteria on a basal enrichment medium (BEM) or tryptose soy agar (TSA) medium supplemented with either 5 mM NP...

  12. Milk protein enriched beverage reduces post-exercise energy intakes in women with higher levels of cognitive dietary restraint

    NARCIS (Netherlands)

    Virgilio, Nicolina; Donno, De Roberta; Bandini, Enrica; Napolitano, Aurora; Fogliano, Vincenzo; Vitaglione, Paola

    2017-01-01

    Objective: The aim of this study was to assess the satiating efficacy of milk proteins compared to carbohydrates in twenty women during post-exercise period. Methods: A milk protein-enriched beverage (MPB), and an isocaloric carbohydrate-enriched beverage (CB) containing respectively 9.3. g and 0.3.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

    Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

    2005-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  16. Solid Fuel - Oxygen Fired Combustion for Production of Nodular Reduced Iron to Reduce CO2 Emissions and Improve Energy Efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Donald R. Fosnacht; Richard F. Kiesel; David W. Hendrickson; David J. Englund; Iwao Iwasaki; Rodney L. Bleifuss; Mathew A. Mlinar

    2011-12-22

    The current trend in the steel industry is an increase in iron and steel produced in electric arc furnaces (EAF) and a gradual decline in conventional steelmaking from taconite pellets in blast furnaces. In order to expand the opportunities for the existing iron ore mines beyond their blast furnace customer base, a new material is needed to satisfy the market demands of the emerging steel industry while utilizing the existing infrastructure and materials handling capabilities. This demand creates opportunity to convert iron ore or other iron bearing materials to Nodular Reduced Iron (NRI) in a recently designed Linear Hearth Furnace (LHF). NRI is a metallized iron product containing 98.5 to 96.0% iron and 2.5 to 4% C. It is essentially a scrap substitute with little impurity that can be utilized in a variety of steelmaking processes, especially the electric arc furnace. The objective of this project was to focus on reducing the greenhouse gas emissions (GHG) through reducing the energy intensity using specialized combustion systems, increasing production and the use of biomass derived carbon sources in this process. This research examined the use of a solid fuel-oxygen fired combustion system and compared the results from this system with both oxygen-fuel and air-fuel combustion systems. The solid pulverized fuels tested included various coals and a bio-coal produced from woody biomass in a specially constructed pilot scale torrefaction reactor at the Coleraine Minerals Research Laboratory (CMRL). In addition to combustion, the application of bio-coal was also tested as a means to produce a reducing atmosphere during key points in the fusion process, and as a reducing agent for ore conversion to metallic iron to capture the advantage of its inherent reduced carbon footprint. The results from this study indicate that the approaches taken can reduce both greenhouse gas emissions and the associated energy intensity with the Linear Hearth Furnace process for converting

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

    International Nuclear Information System (INIS)

    Neff, T.L.

    1982-03-01

    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

  18. Enriched Housing Reduces Disease Susceptibility to Co-Infection with Porcine Reproductive and Respiratory Virus (PRRSV and Actinobacillus pleuropneumoniae (A. pleuropneumoniae in Young Pigs.

    Directory of Open Access Journals (Sweden)

    Ingrid D E van Dixhoorn

    Full Text Available Until today, anti-microbial drugs have been the therapy of choice to combat bacterial diseases. Resistance against antibiotics is of growing concern in man and animals. Stress, caused by demanding environmental conditions, can reduce immune protection in the host, influencing the onset and outcome of infectious diseases. Therefore psychoneuro-immunological intervention may prove to be a successful approach to diminish the impact of diseases and antibiotics use. This study was designed to investigate the effect of social and environmental enrichment on the impact of disease, referred to as "disease susceptibility", in pigs using a co-infection model of PRRSV and A. pleuropneumoniae. Twenty-eight pigs were raised in four pens under barren conditions and twenty-eight other pigs were raised in four pens under enriched conditions. In the enriched pens a combination of established social and environmental enrichment factors were introduced. Two pens of the barren (BH and two pens of the enriched housed (EH pigs were infected with PRRSV followed by A. pleuropneumoniae, the other two pens in each housing treatment served as control groups. We tested if differences in disease susceptibility in terms of pathological and clinical outcome were related to the different housing regimes and if this was reflected in differences in behavioural and immunological states of the animals. Enriched housed pigs showed a faster clearance of viral PRRSV RNA in blood serum (p = 0.014 and histologically 2.8 fold less interstitial pneumonia signs in the lungs (p = 0.014. More barren housed than enriched housed pigs developed lesions in the lungs (OR = 19.2, p = 0.048 and the lesions in the barren housed pigs showed a higher total pathologic tissue damage score (p<0.001 than those in enriched housed pigs. EH pigs showed less stress-related behaviour and differed immunologically and clinically from BH pigs. We conclude that enriched housing management reduces disease

  19. Feeding the nuclear fuel cycle with a long term view; AREVA's front-end business units, uranium mining, UF6 conversion and isotopic enrichment

    International Nuclear Information System (INIS)

    Capus, G.A.P.; Autegert, R.

    2005-01-01

    As a leading provider of technological solutions for nuclear power generation and electricity transmission, the AREVA group has the unique capability of offering a fully integrated fuel supply, when requested by its customers. At the core of the AREVA group, COGEMA Front End Division is an essential part of the overall fuel supply chain. Composed of three Business Units and gathering several subsidiaries and joint 'ventures, this division enjoys several leading positions as shown by its market shares and historical production records. Current Uranium market evolutions put the natural uranium supply under focus. The uranium conversion segment also recently revealed some concerning evolutions. And no doubt, the market pressure will soon be directed also at the enrichment segment. Looking towards the long term, AREVA strongly believes that a nuclear power renewal is needed, especially to help limiting green house effect gas release. Therefore, to address future supplies needed to fuel the existing fleet of nuclear power plants, but also new ones, the AREVA group is planning very significant investments to build new facilities in all the three front-end market segments. As far as uranium mining is concerned, these new mines will be based upon uranium reserves of outstanding quality. As for uranium conversion and enrichment, two large projects will be based on the most advanced technologies. This paper is aimed at recalling COGEMA Front End Division experience, the current status of its plants and operating entities and will provide a detailed overview of its major projects. (authors)

  20. Conceptual assessment and thermal hydraulic analysis of MVDS system for the dry storage of reduced metal fuel

    International Nuclear Information System (INIS)

    Lee, J. C.; Bang, K. S.; Shin, H. S.; Joo, J. S.; Su, K. S.; Kim, H. D.

    2003-01-01

    Conceptual assessment and thermal hydraulic analysis of MVDS storage system have been carried out for application of reduced metal fuel. The storage concept was established considering the optimum weight, storage volume and thermal efficiency. The capacity of MVDS system for loading the reduced metal fuel has four times as compared with existing PWR fuel storage system. In the results of thermal analysis, the maximum temperature of metal fuel was estimated to be 110 .deg. C which is lower than the allowable value under normal operation condition. Therefore, it is shown that the MVDS system can feasibly accomodate the reduced metal fuel in aspect of thermal safety

  1. 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

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    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 {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs 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.

  2. 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.

  3. Comparison of control rod effectiveness for thorium and low-enriched fuel cycles in the GA-1, 160 MW(e) design

    Energy Technology Data Exchange (ETDEWEB)

    Neef, Hans Joachim

    1974-03-15

    In an investigation of the properties of the Thorium-Uranium (Th) and the Low-Enriched Uranium (LEU) fuel cycles it is also necessary to compare the effectiveness of the control rods in a reactor system operating with these sorts of fuel. Furthermore, it is under consideration to start a reactor with LEU fuel and switch-over to a Th cycle. It is also of interest to look at the switch-over phase in respect to the control rod effectiveness. The various fuel cycles have been studied for the same fuel element and control rod design, namely the one of GA's commercially available 1,160 MW(e) reference power station. This paper gives the first results on the control rod calculations and is presented mainly in two parts. Part 1 describes spectral effects which have been investigated by cell calculations with a discrete ordinates transport code. The main result is the higher effectiveness of a rod in a Th-cycle compared with a LEU-cycle. Part 2 reports on reactor calculations with a diffusion code and shows that this advantage can partially disappear in the reactor because of the spatial flux distribution. This effect has to be studied in further investigations for a full understanding.

  4. 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 

  5. Quality improvement of biodiesel blends using different promising fuel additives to reduce fuel consumption and NO emission from CI engine

    International Nuclear Information System (INIS)

    Imdadul, H.K.; Rashed, M.M.; Shahin, M.M.; Masjuki, H.H.; Kalam, M.A.; Kamruzzaman, M.; Rashedul, H.K.

    2017-01-01

    Highlights: • Pentanol, EHN and DTBP are promising fuel additives for improving properties of biodiesel blends. • The utilization of additives improved the properties such as the cetane number, viscosity and oxidation stability. • BSFC, NO and smoke of the EHN and DTBP treated blends are improved by the addition of fuel additives. • Cylinder pressure and Heat Release Rate are enhanced with EHN and DTBP addition. - Abstract: Considering the low cetane number of biodiesel blends and alcohols, ignition promoter additives 2-ethylhexyl nitrate (EHN) and di-tertiary-butyl peroxide (DTBP) was used in this study at a proportion of 1000 and 2000 ppm to diesel-biodiesel-pentanol blends. Five carbon pentanol was used at a proportion of 10% with 20% jatropha biodiesel-70% diesel blends and engine testing was carried out in a single cylinder DI diesel engine. The fuel properties, engine performance, emission and combustion were studied and mainly the effects of two most widely used ignition promoter on the engine behaviour were compared and analyzed. Experimental results indicated that, the fuel properties like density (0.36–1.45%), viscosity (0.26–3.77%), oxidation stability (5.5–26.4%), cetane number (2–14.58%) are improved remarkably with a moderate change in calorific value for the pentanol and ignition promoter treated biodiesel blends depending on the proportion used and for different benchmark. The brake power (BP) is developed very slightly (0.66–1.52%), which is still below than that of diesel, however, the brake specific energy consumption (BSEC) decreased significantly (0.92–5.84%). Although mixing of pentanol increased the nitric oxide (NO) (2.15% than JB20) with reducing the hydrocarbon (HC), carbon monoxide (CO) and smoke, however, the addition of EHN and DTBP reduced the NO (2–4.62%) and smoke (3.45–15.5%) emissions showing higher CO (1.3–9.15%) and HC (5.1–17.87%) emission based on percentage of ignition promoter used. The NO emission

  6. The enrichment secondary market

    International Nuclear Information System (INIS)

    Einbund, D.R.

    1986-01-01

    This paper will addresses two topics: the background to the present status of the enrichment secondary market and the future outlook of the secondary market in enrichment services, and the viability of the nuclear fuel brokerage industry. These two topics are inevitably connected, as most secondary market activity, not only in enrichment but also in natural uranium, has traditionally been conducted with the participation of brokers. Therefore, the author interrelates these topics

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  8. Conductive polymer layers to limit transfer of fuel reactants to catalysts of fuel cells to reduce reactant crossover

    Science.gov (United States)

    Stanis, Ronald J.; Lambert, Timothy N.

    2016-12-06

    An apparatus of an aspect includes a fuel cell catalyst layer. The fuel cell catalyst layer is operable to catalyze a reaction involving a fuel reactant. A fuel cell gas diffusion layer is coupled with the fuel cell catalyst layer. The fuel cell gas diffusion layer includes a porous electrically conductive material. The porous electrically conductive material is operable to allow the fuel reactant to transfer through the fuel cell gas diffusion layer to reach the fuel cell catalyst layer. The porous electrically conductive material is also operable to conduct electrons associated with the reaction through the fuel cell gas diffusion layer. An electrically conductive polymer material is coupled with the fuel cell gas diffusion layer. The electrically conductive polymer material is operable to limit transfer of the fuel reactant to the fuel cell catalyst layer.

  9. Applying burnable poison particles to reduce the reactivity swing in high temperature reactors with batch-wise fuel loading

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Dam, H. van; Hagen, T.H.J.J. van der

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble with a radius of 3 cm containing 9 g of 8% enriched uranium and burnable poison particles (BPP) made of B 4 C highly enriched in 10 B. The radius of the BPP and the number of particles per fuel pebble have been varied to find the flattest reactivity-to-time curve. It was found that for a k∞ of 1.1, a reactivity swing as low as 2% can be obtained when each fuel pebble contains about 1070 BPP with a radius of 75 μm. For coated BPP that consist of a graphite kernel with a radius of 300 μm covered with a B 4 C burnable poison layer, a similar value for the reactivity swing can be obtained. Cylindrical particles seem to perform worse. In general, the modification of the geometry of BPP is an effective means to tailor the reactivity curve of HTRs

  10. The potential role of alcohol fuels in reducing carbon dioxide emissions

    International Nuclear Information System (INIS)

    Duff, S.J.B.

    1991-01-01

    Atmospheric concentrations of CO 2 have increased from 280 to 350 mg/l over the past two hundred years. One of the principal causes has been the increased reliance on combustion of fossil fuels to generate energy. Higher CO 2 levels have been historically correlated with warming of the earth. While attempts have been made to quantify and model the relationships between carbon dioxide emissions, atmospheric CO 2 concentrations, and global climate changes, the state of the current knowledge base is such that large uncertainties persist. It is precisely these uncertainties which has evoked justifiable concern among the scientific community. The use of biomass fuels such as alcohols can provide a partial solution to the problem of increasing emissions of CO 2 . Combustion of biomass fuels releases carbon previously sequestered from the atmosphere during growth. There is a cycling of carbon, with net additions to the atmosphere resulting only from losses, or the use of fossil fuels for process energy. Alcohol fuels can make their biggest impact in the transportation sector, which, in industrial nations, contributes up to 32% of CO 2 emissions. While not the complete answer, alcohol fuels can make a significant impact, and will no doubt be one factor in a multidimensional approach to reducing CO 2 emissions. 17 refs., 4 figs., 10 tabs

  11. OPTIMIZATION OF SPECIFIC FUEL CONSUMPTION OF HYDROGEN IN COMMERCIAL TURBOFANS FOR REDUCING GLOBAL WARMING EFFECTS

    Energy Technology Data Exchange (ETDEWEB)

    T. Hikmet Karakoc; Onder Turan [School of Civil Aviation, Anadolu University, Eskisehir (Turkey)

    2008-09-30

    The main objective of the present study is to perform minimizing specific fuel consumption of a non afterburning high bypass turbofan engine with separate exhaust streams and unmixed flow for reducing global effect. The values of engine design parameters are optimized for maintaining minimum specific fuel consumption of high bypass turbofan engine under different flight conditions, different fuel types and design criteria. The backbones of optimization approach consisted of elitism-based genetic algorithm coupled with real parametric cycle analysis of a turbofan engine. For solving optimization problem a new software program is developed in MATLAB programming language, while objective function is determined for minimizing the specific fuel consumption. The input variables included the compressor pressure ratio ({pi}{sub c}), bypass ratio ({alpha}) and the fuel heating value [h{sub PR}-(kJ/kg)]. Hydrogen was selected as fuel type in real parametric cycle analysis of commercial turbofans. It may be concluded that the software program developed can successfully solve optimization problems at 10{le}{pi}{sub c}{le}20, 2{le}{alpha}{le}10 and h{sub PR} 120,000 with aircraft flight Mach number {le}0.8.

  12. Enriched Housing Reduces Disease Susceptibility to Co-Infection with Porcine Reproductive and Respiratory Virus (PRRSV) and Actinobacillus pleuropneumoniae (A. pleuropneumoniae) in Young Pigs.

    Science.gov (United States)

    van Dixhoorn, Ingrid D E; Reimert, Inonge; Middelkoop, Jenny; Bolhuis, J Elizabeth; Wisselink, Henk J; Groot Koerkamp, Peter W G; Kemp, Bas; Stockhofe-Zurwieden, Norbert

    2016-01-01

    Until today, anti-microbial drugs have been the therapy of choice to combat bacterial diseases. Resistance against antibiotics is of growing concern in man and animals. Stress, caused by demanding environmental conditions, can reduce immune protection in the host, influencing the onset and outcome of infectious diseases. Therefore psychoneuro-immunological intervention may prove to be a successful approach to diminish the impact of diseases and antibiotics use. This study was designed to investigate the effect of social and environmental enrichment on the impact of disease, referred to as "disease susceptibility", in pigs using a co-infection model of PRRSV and A. pleuropneumoniae. Twenty-eight pigs were raised in four pens under barren conditions and twenty-eight other pigs were raised in four pens under enriched conditions. In the enriched pens a combination of established social and environmental enrichment factors were introduced. Two pens of the barren (BH) and two pens of the enriched housed (EH) pigs were infected with PRRSV followed by A. pleuropneumoniae, the other two pens in each housing treatment served as control groups. We tested if differences in disease susceptibility in terms of pathological and clinical outcome were related to the different housing regimes and if this was reflected in differences in behavioural and immunological states of the animals. Enriched housed pigs showed a faster clearance of viral PRRSV RNA in blood serum (p = 0.014) and histologically 2.8 fold less interstitial pneumonia signs in the lungs (p = 0.014). More barren housed than enriched housed pigs developed lesions in the lungs (OR = 19.2, p = 0.048) and the lesions in the barren housed pigs showed a higher total pathologic tissue damage score (ppigs. EH pigs showed less stress-related behaviour and differed immunologically and clinically from BH pigs. We conclude that enriched housing management reduces disease susceptibility to co-infection of PRRSV and A

  13. Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2017-01-01

    Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

  14. Advanced Neutron Source enrichment study

    International Nuclear Information System (INIS)

    Bari, R.A.; Ludewig, H.; Weeks, J.R.

    1996-01-01

    A study has been performed of the impact on performance of using low-enriched uranium (20% 235 U) or medium-enriched uranium (35% 235 U) as an alternative fuel for the Advanced Neutron Source, which was initially designed to use uranium enriched to 93% 235 U. 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

  15. Development of Nuclear Renewable Oil Shale Systems for Flexible Electricity and Reduced Fossil Fuel Emissions

    Energy Technology Data Exchange (ETDEWEB)

    Daniel Curtis; Charles Forsberg; Humberto Garcia

    2015-05-01

    We propose the development of Nuclear Renewable Oil Shale Systems (NROSS) in northern Europe, China, and the western United States to provide large supplies of flexible, dispatchable, very-low-carbon electricity and fossil fuel production with reduced CO2 emissions. NROSS are a class of large hybrid energy systems in which base-load nuclear reactors provide the primary energy used to produce shale oil from kerogen deposits and simultaneously provide flexible, dispatchable, very-low-carbon electricity to the grid. Kerogen is solid organic matter trapped in sedimentary shale, and large reserves of this resource, called oil shale, are found in northern Europe, China, and the western United States. NROSS couples electricity generation and transportation fuel production in a single operation, reduces lifecycle carbon emissions from the fuel produced, improves revenue for the nuclear plant, and enables a major shift toward a very-low-carbon electricity grid. NROSS will require a significant development effort in the United States, where kerogen resources have never been developed on a large scale. In Europe, however, nuclear plants have been used for process heat delivery (district heating), and kerogen use is familiar in certain countries. Europe, China, and the United States all have the opportunity to use large scale NROSS development to enable major growth in renewable generation and either substantially reduce or eliminate their dependence on foreign fossil fuel supplies, accelerating their transitions to cleaner, more efficient, and more reliable energy systems.

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

    International Nuclear Information System (INIS)

    Fishbone, L.G.; Moussalli, G.; Naegele, G.; Ikonomou, P.; Hosoya, M.; Scott, P.; Fager, J.; Sanders, C.; Colwell, D.; Joyner, C.J.

    1994-01-01

    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. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. 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. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant

  17. The plutonium fuel cycles

    International Nuclear Information System (INIS)

    Pigford, T.H.; Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

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

    KAUST Repository

    Yan, Hengjing; Saito, Tomonori; Regan, John M.

    2012-01-01

    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

  19. Promotion of uranium enrichment business

    International Nuclear Information System (INIS)

    Kurushima, Morihiro

    1981-01-01

    The Committee on Nuclear Power has studied on the basic nuclear power policy, establishing its five subcommittees, entrusted by the Ministry of Nternational Trade and Industry. The results of examination by the subcommittee on uranium enrichment business are given along with a report in this connection by the Committee. In order to establish the nuclear fuel cycle, the aspect of uranium enrichment is essential. The uranium enrichment by centrifugal process has proceeded steadily in Power Reactor and Nuclear Fuel Development Corporation. The following matters are described: the need for domestic uranium enrichment, the outlook for overseas enrichment services and the schedule for establishing domestic enrichment business, the current state of technology development, the position of the prototype enrichment plant, the course to be taken to establish enrichment business the main organization operating the prototype and commercial plants, the system of supplying centrifuges, the domestic conversion of natural uranium the subsidies for uranium enrichment business. (J.P.N.)

  20. Reduced Gravity Studies of Soret Transport Effects in Liquid Fuel Combustion

    Science.gov (United States)

    Shaw, Benjamin D.

    2004-01-01

    Soret transport, which is mass transport driven by thermal gradients, can be important in practical flames as well as laboratory flames by influencing transport of low molecular weight species (e.g., monatomic and diatomic hydrogen). In addition, gas-phase Soret transport of high molecular weight fuel species that are present in practical liquid fuels (e.g., octane or methanol) can be significant in practical flames (Rosner et al., 2000; Dakhlia et al., 2002) and in high pressure droplet evaporation (Curtis and Farrell, 1992), and it has also been shown that Soret transport effects can be important in determining oxygen diffusion rates in certain classes of microgravity droplet combustion experiments (Aharon and Shaw, 1998). It is thus useful to obtain information on flames under conditions where Soret effects can be clearly observed. This research is concerned with investigating effects of Soret transport on combustion of liquid fuels, in particular liquid fuel droplets. Reduced-gravity is employed to provide an ideal (spherically-symmetrical) experimental model with which to investigate effects of Soret transport on combustion. The research will involve performing reduced-gravity experiments on combustion of liquid fuel droplets in environments where Soret effects significantly influence transport of fuel and oxygen to flame zones. Experiments will also be performed where Soret effects are not expected to be important. Droplets initially in the 0.5 to 1 mm size range will be burned. Data will be obtained on influences of Soret transport on combustion characteristics (e.g., droplet burning rates, droplet lifetimes, gas-phase extinction, and transient flame behaviors) under simplified geometrical conditions that are most amenable to theoretical modeling (i.e., spherical symmetry). The experiments will be compared with existing theoretical models as well as new models that will be developed. Normal gravity experiments will also be performed.

  1. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    International Nuclear Information System (INIS)

    Fukaya, Y.; Okubo, T.; Uchikawa, S.

    2008-01-01

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

  2. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2008-07-15

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

  3. The conceptual design of the standard and the reduced fuel assemblies for an advanced research reactor

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Cho, Yeong Garp; Yoon, Doo Byung; Dan, Ho Jin; Chae, Hee Tack; Park, Cheol

    2005-01-01

    HANARO (Hi-flux Advanced Neutron Application Reactor), is an open-tank-in-pool type research reactor with a thermal power of 30MW. The HANARO has been operating at Korea Atomic Energy Research Institute since 1995. Based on the technical experiences in design and operation for the HANARO, the design of an Advanced Research Reactor (ARR) was launched by KAERI in 2002. The final goal of the project is to develop a new and advanced research reactor model which is superior in safety and economical aspects. This paper summarizes the design improvements of the conceptually designed standard fuel assembly based on the analysis results for the nuclear physics. It includes also the design of the reduced fuel assembly in conjunction with the flow tube as the fuel channel and the guide of the absorber rod. In the near future, the feasibility of the conceptually designed fuel assemblies of the ARR will be verified by investigating the dynamic and the thermal behaviors of the fuel assembly submerged in coolant

  4. Development of a treatment technology for diluting highly enriched AL-based DOE spent nuclear fuel: principles and practices

    International Nuclear Information System (INIS)

    Adams, T.M.; Duncan, A.J.; Peacock, H.B.; Fisher, D.L.

    2001-01-01

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage. Interim storage canisters at SRS will be sent to a Mined Geologic Repositor