WorldWideScience

Sample records for average fuel enrichment

  1. Contemporary and prospective fuel cycles for VVER-440 based on new assemblies with higher Uranium capacity and higher average fuel enrichment

    International Nuclear Information System (INIS)

    RRC Kurchatov Institute has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for VVER-440 reactors. Works were performed to upgrade and improve VVER-440 fuel cycles on the basis of second generation fuel assemblies allowing core thermal power to be up rated 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 VVER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of VVER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (author)

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

    International Nuclear Information System (INIS)

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

  3. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

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

  4. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

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

  5. Low enriched fuels for NRU

    International Nuclear Information System (INIS)

    Low enriched uranium silicide dispersion fuels are under development for use in Canadian test reactors. These 20% enriched dispersion fuels are to replace the current 93% enriched uranium-aluminum alloy driver fuels. The reduced enrichment is intended to reduce the risk of illegal diversion for weapons proliferation. Developments in uranium silicide dispersion manufacturing technology have proven the production viability of the fuel. Success in the irradiation testing of the dispersion fuels in mini-element form has led to the irradiation of seven full-size fuel assemblies

  6. 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.

  7. Enrichment of solid fuels. Proceedings

    International Nuclear Information System (INIS)

    The papers deal with the methods for solid fuels enrichment, spatially coals, biomass and solid wastes, in order to improve their quality, as well as the price and interest of their use. The priority is given to the Macedonian lignite due to its low calorific value, moisture, ash content and bad mechanical characteristics. The holding of this Meeting was initiated by ZEMAK - Association of power engineers of Macedonia. Papers relevant to INIS are indexed separately

  8. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    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.)

  9. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    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

  10. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

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

  11. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

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

  12. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

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

  13. Fuel performance experience with slightly-enriched uranium

    International Nuclear Information System (INIS)

    There is an economic incentive associated with an increase in the average burnup of fuel discharged from CANDU reactors. The slightly enriched uranium fuel cycle provides a path by which increased fuel burnup can be attained in the short term. This paper discusses the effects upon fuel performance of increases in burnup, the existing base of pertinent data and the additional work that is required to verify the technology. Areas of fuel performance that are of particular importance at high burnup are fission gas release and power ramp behaviour. Fuel performance models will also require development and verification. The development program is in two phases, with the first phase leading to intermediate burnups and the second phase providing higher burnups

  14. Step II fuel, many kinds of enrichment initial charge core

    International Nuclear Information System (INIS)

    In the design of the core and the fuel of BWRs, while the heightening of burnup has been advanced stepwise, and the reliability of fuel has been ensured, the reduction of spent fuel and the heightening of operation performance have been promoted. At present, high burnup 8 x 8 step 2 fuel of taking-out average burnup 39.5 GWd/t is practically used. Moreover, toward higher burnup, the step 3 fuel of 9 x 9 arrangement and taking-out average burnup 45 GWd/t is developed and prepared. The technology of heightening the burnup of BWR fuel is applicable also to the replacement fuel for existing BWRs. No. 3 plant in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. (1100 MWe output, BWR5 type) started the commercial operation in August, 1993. As for the design of the core and the fuel of this plant, the core design of many kinds of enrichment initial charge using high burnup 8 x 8 fuel was adopted. The care design of from No. 1 to No. 5 plants in Kashiwazaki Kariwa Nuclear Power Station followed the history of the improved core design for initial charge of BWRs. As the way of thinking on initial charge core design, the optimization of the technology of uranium saving, the increase of degree of enrichment, the adoption of 8 x 8 fuel and long life control rods are explained. The design of high burnup 8 x 8 fuel and the results of start-up operation are reported. (K.I.)

  15. Fuel cell operation with oxygen enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Fournier, M.; Hamelin, J.; Agbossou, K.; Bose, T.K. [Universite du Quebec a Trois-Rivieres, Institut de Recherche sur l' Hydrogene, 3351, Boul. Des Forges, C.P. 500, Trois-Rivieres (QC), G9A 5H7 (Canada)

    2003-02-01

    Experimental results on the performance of a Ballard 5 kW proton exchange membrane fuel cell stack for different oxygen contents in the oxidant are presented. A description of the experimental setup is given. Polarization, power, and efficiency curves as a function of the current density, for different oxygen concentrations are presented. This detailed characterization of the fuel cell stack behavior is required in order to evaluate the effects of oxygen enrichment on the net power output of the stack. This investigation is done in the framework of a project on stand-alone power generation systems using renewable energy sources, and based on hydrogen production and storage. An electrolyzer, powered by the excess electrical energy from renewable energy sources, produces hydrogen. The stored hydrogen could then be used to feed an energy conversion device, such as a fuel cell stack, which acts as a secondary power source in periods of high demand. Therefore, a second objective is to evaluate the possibility of using the oxygen produced by the electrolyzer for the enrichment. Other oxygen enrichment techniques such as membrane gas separation and pressure swing adsorption are also discussed. Net available power and system efficiency are used as comparison factors. (Abstract Copyright [2002], Wiley Periodicals, Inc.)

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

    International Nuclear Information System (INIS)

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

  17. Thermal hydraulic calculations for TNRR using low enriched fuel

    International Nuclear Information System (INIS)

    This paper presents the preliminary results of the thermal-hydraulic calculations of the Tajoura nuclear research reactor (TNRR)) using low enrichment fuel 36%. The study considered the fresh core compact load. With the assumptions used in the calculation, no significant change in the thermal hydraulic parameters was noticed. A comparison was made for the most heated element between the (80% U235) enriched fuel and the (36% U235) enriched fuel under normal operating conditions. a slight increase in the maximum fuel surface temperature was noticed. These changes were due to change in the fuel material which is oxide fuel and to the increase in the meat thickness. (author)

  18. Development of long-life low enrichment fuel

    International Nuclear Information System (INIS)

    With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on non-proliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U. S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of this year, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  19. Low-enriched fuel particle performance review. [UO2

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

  1. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Science.gov (United States)

    2010-10-01

    ...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not exist... section for which a fuel economy value has been neither determined nor approved under 40 CFR part 600, the... subpart F of 40 CFR part 600. (c) Average fuel economy. Average fuel economy must be based upon...

  2. RERTR program progress in qualifying reduced-enrichment fuels

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J.L.

    1982-01-01

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

  3. Delays hit conversion to low enriched uranium fuel

    Science.gov (United States)

    Allen, Michael

    2016-03-01

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

  4. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

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

  5. Development of low enrichment MTR fuel at Dounreay

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  7. 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.

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

    Science.gov (United States)

    Tarzi, Zahi; Speyer, Jason; Wirz, Richard

    2013-05-01

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

  9. A NDT Method For Uranium Fuel Enrichment Verification

    International Nuclear Information System (INIS)

    Nondestructive testing of fresh uranium fuel materials as a part of a QA programme in a nuclear industry is wide accepted today. These inspections of the fuel materials are based on the well known methods. As a participation of the NET Laboratory in the new QA programme at the VINCA Institute, a method for the absolute determination of the enrichment value of fresh high enriched uranium fuel segments, based on a gamma-ray spectroscopy, is developed. It is primary based on the application of a developed ANA computer code for a gamma spectrum analysis, and the initial experience gained in the previous works on the QA at the VINCA Institute. (author)

  10. Challenging cycles of Dukovany NPP with highly enriched fuel optimized by the Athena code

    International Nuclear Information System (INIS)

    The Dukovany NPP has operated four WWER-440 reactor units. Starting 2009 power of third unit has been up rated, so reactor is operated at power level of 1444 MWt instead of 1375 MWt used up to now. Innovation of secondary circuit and electrical equipment leads to electrical output of 500 MWe. During four years all four units will be up rated this way. Power up rates together with outage shortages are challenges for fuel cycles optimization. Reactors of Dukovany NPP have already operated in incomplete five-year fuel cycles based on modified Gd-2+ fuel design called Gd-2M with 4.38 % average enrichment now. New fuel types with high enrichment were proposed by fuel manufacture (4.87 % average enrichment) and by SKODA JS Company (4.76 % average enrichment). These fuels should allow complete five year fuel cycle with higher power and has also ability to allow incomplete six-year fuel cycle. Fuels should economize fuel assemblies and operational costs. The SKODA JS Company together with the University of West Bohemia has developed new optimizing code called Athena. The code has been used to optimize fuel cycles from twenty fifth cycle of third unit of Dukovany NPP up to thirty fourth cycle with new fuel types presented above. Comparison between fuel cycles with Gd-2M fuel and with mixed loading patterns (Gd-2M mixed with new types) has been calculated. Positives and negatives are discussed from physical, safety, operational, and economical points of view. New fuels allow 330 to 340 FPD long cycles with up rated power, but are very challenging to radial power distribution parameter (FdH). Results show, that operational parameter limit of FdH used until now should be increased from 1.54 to 1.58. Couple of fuel assemblies is economized. Results are a part of study of applicability and feasibility of new fuels and might be important for additional development of fuel design. Results will also be used as a comparison of presently used optimization code Optimal and new Athena

  11. Reduced enriched fuel status at CERCA

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 UO2 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

  14. Evaluation of the reactivity feedback coefficients of the Da Lat Nuclear Research Reactor using highly enriched uranium fuels and low enriched uranium fuels

    International Nuclear Information System (INIS)

    This article presents the calculation results of fuel and moderator temperature coefficients of reactivity for the Da Lat nuclear research reactor (DNRR), using Highly Enriched Uranium (HEU) fuels (36%) and Low Enriched Uranium (LEU) fuels (19.75%). This study uses the WIMSD code to calculate the cross sections of all the reactor components at different temperatures and these group constants were used then in the CITATION code to calculate the effective multiplication factor at distinct moderator and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. The results are the average of fuel temperature coefficient of reactivity for LEU: -2.15(x10-5 ∆k/k/oC), HEU: -1.91(x10-5∆k/k/oC); the average of the moderator temperature coefficients of reactivity for LEU: -2.44(x10-4 ∆k/k/oC),HEU: -2.67(x10-4 ∆k/k/oC). The calculated feedback coefficients were compared with the measured data from DNRR. Good agreements were obtained. Moreover,getting the trend of these coefficients versus the rise in temperature. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Core configuration of the Syrian reduced enrichment fuel MNSR

    International Nuclear Information System (INIS)

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

  19. Isotope correlation studies relative to high enrichment test reactor fuels

    International Nuclear Information System (INIS)

    Several correlations of fission product isotopic ratios with atom percent fission and neutron flux, for highly enriched 235U fuel irradiated in two different water moderated thermal reactors, have been evaluated. In general, excellent correlations were indicated for samples irradiated in the same neutron spectrum; however, significant differences in the correlations were noted with the change in neutron spectrum. For highly enriched 235U fuel, the correlation of the isotopic ratio 143Nd/145+146Nd with atom percent fission has wider applicability than the other fission product isotopic ratio evaluated. The 137Cs/135Cs atom ratio shows promise for correlation with neutron flux. Correlations involving heavy element ratios are very sensitive to the neutron spectrum

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

    International Nuclear Information System (INIS)

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

  1. LVR-15 reactor performance and transformation to low enriched fuel

    International Nuclear Information System (INIS)

    Experimental research reactor LVR-15 situated in Nuclear Research Institute Rez, plc. has been utilized since 1957. The present reactor nominal power is 10 MW. Standard reactor cycle is 21 days and the reactor operates 8 - 10 cycles per year. The State Office for Nuclear Safety newly licensed the reactor till 2014. The reactor is of multi-purpose use. The basic research is carried out using horizontal neutron beams and one of them is used for development and application of the boron neutron captures therapy for brain tumors. Mostly material testing of PWR and BWR specimens is performed in high-pressure loops and irradiation rigs operated in the reactor. Several vertical channels serve for production of neutron transmutation doped silicon and isotopes production for medical purposes. Reactor was originally designed with EK-10 fuel type and consequently reconstructed for HEU with the 80% 235U enrichment in 1987-89. In period 1987-1989 the reactor undertook the second reconstruction to enhance its experimental and commercial utilizing. Later, the fuel enrichment was changed to 36 % with the use of the IRT-2M fuel type. The neutron-physical characteristics of the reactor core and fuel cycle design and analyses are carried out using the WIMS-D4m, NODER and OMEGA programs. The codes were used for a preliminary evaluation of essential changes of main neutronic characteristics of the LVR-15 core with the prospective conversion to low enriched fuel (LEU). Three types of FA-s has been assessed: 1) currently used IRT-2M(36%), 2) IRT-3M(20%), 3) IRT-4M(20%) and results are presented in the article. (author)

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

    International Nuclear Information System (INIS)

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

  3. Development of lower enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

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

  4. Materials Control in the Fabrication of Enriched Uranium Fuels

    International Nuclear Information System (INIS)

    Intense activity in the field of fuel element technology at Oak Ridge National Laboratory during the past 15 years has led to the establishment of sound process and enriched material control procedures that find wide applicability in the commercial fabrication of fuel elements today. Reliable techniques for handling enriched fuel in alloy, dispersion and bulk oxide form were developed and adopted as standards in the course of design and fabrication of prototypic fuel elements for start-up operation of the MTR, Bulk Shielding or ''Swimming Pool'' Reactor, Army Package Power Reactor, Tower Shielding Reactor, Geneva Conference Display Reactor, High Flux Isotope Reactor, and the EGCR. The experience gained serves as background for this paper, which will stress material control problems and their solution during the fabrication of various types of enriched uranium fuel components. The basic objective to be met in the design of a good materials control system are: (1) minimizing the number of material units to be accounted for; (2) designing separate records for each major fabrication step and linking these in a manner that permits isolation of differences with a minimum of effort; (3) integrating the maximum number of controls into the minimum number of records to eliminate duplication; and (4) introducing a sufficient number of cross-checks into the system to ensure reliability. In every fabrication programme, successful control was achieved by establishing a unit procedure in the following areas: (1) starting materials in the as-received form; (2) fabrication of components; (3) component processing; and (4) scrap handling. Consolidation of control records into a master summary was helpful in confirming the materials inventory, evaluating the fabrication process, and preparing management reports. Establishment of sampling methods and examination of results indicated that multiple control is necessary to ensure proper fuel content. Mechanical adjustment and density

  5. 40 CFR 600.510-86 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy... Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy §...

  6. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Science.gov (United States)

    2010-10-01

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

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

    International Nuclear Information System (INIS)

    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

  8. 75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

    Science.gov (United States)

    2010-07-29

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

  9. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

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

  10. Criticality evaluation of BWR MOX fuel transport packages using average Pu content

    International Nuclear Information System (INIS)

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by a homogeneous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, COGEMA LOGISTICS has studied a new calculation method based on the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in our approach. With this new method, for the same package reactivity, the Pu-content allowed in the package design approval can be higher. The COGEMA LOGISTICS' new method allows, at the design stage, to optimise the basket, materials or geometry for higher payload, keeping the same reactivity

  11. 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)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-30

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Utilization of risk information for fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    For the purpose of applying risk information to the ex-post regulation after granting license, and of reflecting it to a manual of technical bases specified in the design and construction authorization, a selection method of items relied on for safety (IROFS) for the uranium fuel fabrication and enrichment facilities has prepared based on national laws, international standards, and the course of license permission. It has also prepared a way of improvement of a manual of technical bases to support the design and construction authorization for the uranium facilities. Furthermore, a tentative evaluation manual to ISA (Integrated Safety Analysis) results has prepared for reviewing the ISA results of the uranium facilities. The manual was effectively used for reviewing the ISA results obtained from the licensee's voluntary field implementation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-07-01

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

  19. The shipment of spent highly enriched fuel elements

    International Nuclear Information System (INIS)

    Along the last quarter of 2000, in the Radioactive Waste Management Area of the Ezeiza Atomic Center, were carried out all the duties associated with the exportation to the United States of America of spent fuel containing high enrichment uranium produced in such country (in the frame of the program for the restitution of this type of fuel, developed by his Department of Energy) that were utilized until 1999 in the operation of the Research and Radioisotope Production Reactor RA3. All the relevant phases of the project are described, including the cropping activities for the conditioning of the items, the canning of them when become necessary, the transferences from the working pool to the transport casks and the final preparations for the shipment. These transport casks were specially designed and constructed with this purpose and licensed by the Nuclear Regulatory Commission of the United States and for the Nuclear Regulatory Authority of Argentina. The transport was done following national and international rules. (author)

  20. 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

    International Nuclear Information System (INIS)

    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

  1. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Science.gov (United States)

    2010-07-01

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

  2. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR part... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel...

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

    International Nuclear Information System (INIS)

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

  4. 76 FR 67765 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding...

    Science.gov (United States)

    2011-11-02

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

  5. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

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

  6. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

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

  7. Canflex: A fuel bundle to facilitate the use of enrichment and fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    The neutron economy of the CANDU reactor results in it being an ideal host for a number of resource-conserving fuel cycles, as well as a number of potential ''symbiotic'' fuel cycles, in which fuel discharged from light-water cooled reactors is recycled to extract the maximum energy from the residual fissile material before it is sent for disposal. The resource conserving fuel cycles include the natural-uranium, slightly-enriched-uranium and thorium fuel cycles. The ''LWR-symbiotic'' cycles include recovered uranium and various options for the direct use of spent LWR fuel in CANDU reactors. However, to achieve the maximum economic potential of these fuel-cycle options requires irradiation to burnups higher than that possible with natural uranium. To provide a basis for the economic use of these fuel cycles, a program is underway to develop and demonstrate a CANDU fuel bundle capable of both higher burnups and greater operating margins. This new bundle design is being developed jointly by AECL and KAERI, and uses smaller-diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This allows operation to burnups greater than 21 MWd/KgU. A combination of this lower peak-element rating, plus development work underway at AECL to enhance the thermalhydraulic characteristics of the bundle (including both critical heat flux and bundle pressure drop), provides a greater operating margin for the bundle. This new bundle design is called CANFLEX, and the program for its development in Canada and Korea is described in this paper. (author). 19 refs, 5 figs

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

    2010-02-01

    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.

  10. The Lessons Learned from Conversion of 10 Mw Research Reactor to Low Enrichment Fuel

    International Nuclear Information System (INIS)

    The conversion of the 10 Mw research reactor of the Institute of Nuclear Physics in Tashkent from 90% enriched fuel to the low (19.7% enrichment) one was done in two major steps. First step, which started in August 1998 and completed in February 1999, was including a conversion of reactor to 36% enrichment fuel. The second stage (started in February 2008 and accomplished in November 2009) allowed the full conversion of reactor to 19.7% enrichment fuel. In parallel with these activities the continuous works on testing of the new types of fuel elements were carried out. We present the review of our experience obtained from conversion of research reactor to the low enriched fuel including theoretical estimates, the chosen geometry of core elements, the results on neutron flux measurements as well as other data on reactor utilization which were accumulated after full conversion of reactor. (author)

  11. The Lessons Learned from Conversion of 10 Mw Research Reactor to Low Enrichment Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuldashev, B. [Stanford University, Stanford, CA 94305-6165 (United States); Baytelesov, S.; Dosimbaev, A.; Kungurov, F.; Maksumov, N.; Salikhbaev, U. [Institute of Nuclear Physics, Uzbekistan Academy of Sciences, Ulugbek, 100132, Tashkent (Uzbekistan)

    2011-07-01

    The conversion of the 10 Mw research reactor of the Institute of Nuclear Physics in Tashkent from 90% enriched fuel to the low (19.7% enrichment) one was done in two major steps. First step, which started in August 1998 and completed in February 1999, was including a conversion of reactor to 36% enrichment fuel. The second stage (started in February 2008 and accomplished in November 2009) allowed the full conversion of reactor to 19.7% enrichment fuel. In parallel with these activities the continuous works on testing of the new types of fuel elements were carried out. We present the review of our experience obtained from conversion of research reactor to the low enriched fuel including theoretical estimates, the chosen geometry of core elements, the results on neutron flux measurements as well as other data on reactor utilization which were accumulated after full conversion of reactor. (author)

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

    International Nuclear Information System (INIS)

    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 UO2 and U-9%Mo fuel with U235 19.7% enrichment are described. (author)

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

    International Nuclear Information System (INIS)

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

  14. Aspects of the contamination with oxygen in obtaining low enriched uranium fuel

    International Nuclear Information System (INIS)

    The manufacturing of TRIGA fuel rods with low enriched uranium follows in principle the same route as high-enriched uranium. The high purity of the primary metals (uranium, zirconium and erbium) is important for determining the equilibrium metal-hydrogen phases. The impurities from the metal, on the surface and from hydrogen may have an important influence on the hydriding process. This paper presents the aspects of the fuel contamination with oxygen during the manufacturing process of the low enriched uranium fuel. The continuous control of the oxygen concentration in the working zone ensures avoidance of the accidental contamination. Key words: manufacturing, fuel, oxygen, contamination. (authors)

  15. Consideration of methods to determine an enrichment of commercial fast reactor fuel

    International Nuclear Information System (INIS)

    In the situation that a variety of fuel composition is fed to fast breeder reactors (FBRs) during LWR-to-FBR transition stage, a special consideration on fuel reactivity and its loss by burnup is needed for each feeding fuel to keep criticality during operation period as prescribed. This paper describes some characteristics of the methods used to determine an enrichment of fast reactor fuel from the core design points of view. 5 methods have been considered: 1) Align reactivity worth of fresh fuel, 2) Align reactivity worth of 1 cycle-burnt fuel, 3) Align reactivity worth of discharged fuel, 4) Align reactivity worth of fuel at the end of equilibrium cycle, and 5) Ad-hoc determination of fuel enrichment keeping criticality cycle-by-cycle. Some other methods may be possible. However, they would be nothing more than the minor changes of 1) to 5). The characteristics of the 5 methods to determine an enrichment of fast reactor fuel have been evaluated from the viewpoints of core design by 2-dimensional neutronic calculation. Enrichments of fuels and various core characteristics such as the excess reactivity, the maximum linear power, and the total maximum assembly power have been investigated by using two kinds of typical fuel composition representing LWR spent fuel with short and long cooling times. Merits and demerits of these methods are clarified

  16. ZPR-3 Assembly 12: A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average 235U enrichment of 21 atom %

    International Nuclear Information System (INIS)

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was 235U or 239Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core 235U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically

  17. ZPR-3 Assembly 11: A cylindrical sssembly of highly enriched uranium and depleted uranium with an average 235U enrichment of 12 atom % and a depleted uranium reflector

    International Nuclear Information System (INIS)

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was 235U or 239Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core 235U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark

  18. 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)

    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% 235U 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 would

  19. Medium-enriched uranium/thorium fuel cycle parametric studies for the HTGR

    International Nuclear Information System (INIS)

    Operation of HTGRs on proliferation-resistant medium-enriched uranium/thorium fuel cycles is feasible based on the findings of fuel cycle parametric studies conducted for the Department of Energy by General Atomic Company. The analyses performed to evaluate the feasibility and optimization of such fuel cycles are described. Primary variables considered in arriving at optimum designs included cycle length, fuel particle and fuel rod dimensions, the carbon-to-thorium ratio, and the refueling frequency

  20. Final qualification testing results for TRIGA low-enriched uranium fuel

    International Nuclear Information System (INIS)

    Following the adoption of policies by the US government and other fuel supplier nations to limit, with few exceptions, export of fuels to those enriched to 235U, GA Technologies undertook, starting in 1976, a rigorous program to develop uranium zirconium-hydride fuels with high uranium concentrations up to 3.7 g U/cm3, while limiting the enrichment to 235U. By increasing the uranium concentration from 8.5 to 45 wt% and by including erbium as a burnable poison, the long reactivity lifetime of the high-enriched uranium cores has been preserved. The achieved purpose of these low-enriched uranium (LEU) fuels was also to preserve all the unique safety features of the TRIGA fuel system - prompt negative temperature coefficient of reactivity, high fission product retentivity, chemical stability when quenched from high temperatures in water, and dimensional stability over a broad range of operating temperatures

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

    International Nuclear Information System (INIS)

    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 235U) UAl3 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 UAl3 fuel with high-density, low enriched (235U) 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 U3Si2 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)

  2. Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs

    International Nuclear Information System (INIS)

    Fuel cycle costs with high-enrichment uranium (HEU) and low-enrichment uranium (LEU) fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are an increase of about 3% in HEU fuel cycle costs and a reduction of 2-3% in the fuel cycle costs for LEU silicide fuel

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    (Mechanically assembled) 90% enriched. The work carried out so far can be summarised as follows: Precipitation trials from uranyl nitrate followed by calcining of 800 deg C and further treatment at 1400 deg C have resulted in the production of U3O8 particles in the desired size range of +325/-100 mesh with densities of up to 82 g/cc and surface areas down to 0.06 m2/g. Studies of variables are continuing in order to maximise the production of particles within the above size range; Pressing trials at various pressures have confirmed that compacts averaging 95% theoretical density can be produced without the use of a binder. A study of the use of power-weighing as an assessment of uranium content indicates that a precision of ± 0.5% can be achieved; this is comparable with current instrument measuring techniques; Rolling trials have shown that close control of core dimensions and good metallurgical bonding can be achieved using an aluminium-clad 6082 alloy for the clads and frames. Dogboning, clad thickness and uranium segregation determined by radiography are within the limits currently applied to U/Al alloy plates; Welding and forming trials on the plates produced have been satisfactory and it is not anticipated that difficulties will be encountered in fuel element assembly; At the present time, prototype fuel elements, both concentric and plate type, containing U3O8/Al cermet cores at 45% enrichment are being manufactured for irradiation trials. Proposals for these trials are being studied by the appropriate reactor safety committees and it is anticipated that irradiation in UK reactors will commence early in 1981. Arrangements are being made for post-irradiation examination of the fuel elements

  5. Low enriched aluminide and silicide fuel element technology at B and W (USA)

    International Nuclear Information System (INIS)

    Babcock and Wilcox is fabricating full size fuel elements with low enriched uranium silicide and uranium aluminide. BandW also provides high enrichred U3O8 and UA Lsub(x) for United States Research Reactors, and Test Research and Training Reactors (TRTR). BandW and Argonne National Laboratry (ANL) are actively involved in the Reduced Enrichment Research and Test Reactor (RERTR) Program and have undertaken a joint effort in which BandW is fabricating two Oak Ridge Reactor (ORR ) elements with uranium silicide fuel. During plate development, fuel plates were fabricated with compacts containing U3SiAl and U3Si2 fuel. (author)

  6. Slightly enriched uranium in CANDU: An economic first step towards advanced fuel cycles

    International Nuclear Information System (INIS)

    The natural-uranium fuelled Canada Deuterium-Uranium (CANDU) nuclear reactor system has proven to be a safe, reliable and economical producer of electricity for over a quarter of a century. The CANDU system, however, is not restricted to the use of natural-uranium fuel; a wide range of advanced fuel cycles can be accommodated. In the short term, slightly enriched uranium (SEU) is the most promising of these advanced fuel cycles. SEU offers a reduction in the total fuel cycle cost of between 25 and 50% relative to natural-uranium fuel. Uranium consumption is decreased by 30 to 40%. In addition the volume of spent fuel is reduced by a factor of two to three, depending on the enrichment selected. SEU also offers greater flexibility in the design of future CANDU reactors. A variety of fuel management options can be employed in CANDU with slightly enriched fuels. Fuel performance is expected to be good for the burnups of interest, but further fuel testing is planned and is currently in progress in order to confirm this. Programs in place at Atomic Energy of Canada Limited (AECL) will lead to the demonstration and introduction of slightly enriched uranium in CANDU. Ontario Hydro, a Canadian utility with twenty CANDUs operating or under construction, is considering a program which could lead to the implementation of SEU in its nuclear generating stations. (author). 30 refs, 7 figs

  7. 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)

    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 235U 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 rather

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

    International Nuclear Information System (INIS)

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

  9. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    International Nuclear Information System (INIS)

    The RB reactor was designed as a natural-uranium, heavy water, non reflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments. (author)

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

    International Nuclear Information System (INIS)

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

  11. The ''RB'' reactor uranium fuel enrichment verification by gamma-ray spectroscopy

    International Nuclear Information System (INIS)

    Gamma spectrometry analysis of natural and 2% enriched uranium metal fuel at the RB reactor was performed by germanium gamma spectrometer applying developed computer code ANA. Different samples of the RB reactor uranium fuel, placed at various distances from the Ge detector, were used during measurements. Gamma-ray self-absorption in the fuel material and the geometrical corrections were included in the calculation performed by computer code EFI based on a Monte Carlo method. Evaluated experimental data were used to determine branching ratio for the 1001 keV gamma line of 234mPa which is in equilibrium with 238U. Obtained results were in good agreement with the results of other authors. Applied gamma spectrometry method is used for examination of the fresh fuel composition and validation of isotopic enrichment of the 2% enriched uranium fuel at the RB reactor. (author)

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

    International Nuclear Information System (INIS)

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

  13. Use of fuel elements and fuel rod arrays of WWER-type with 20 % enriched cermet fuel for reactors of floating power plant KLT-40S

    International Nuclear Information System (INIS)

    It was carried out numerical analysis of the physical characteristics of change from normal active zone to fuel elements and fuel rod arrays using fuel cycle of WWER-1000 type as well as at replacement of oxide fuel to cermet fuel (60%UO2+40% of silumin) with 20% enrichment. At that the main physical characteristics of active zone and reactor are kept - geometric sizes, power, coolant properties etc. It was given the main physical properties of fuel elements and fuel rod arrays of active zone with cermet fuel. Calculation of neutron physical characteristics was carried out. The reactor has internal self-protectability

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

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

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

    International Nuclear Information System (INIS)

    Gamma-ray passive technique is a very effective method to assay and determine 235U enrichment of nuclear power plant fuel rods. There is a weakness in this passive method, i.e. only after the uranium isotope daughters of UO2 pellets have reached to equilibrium with uranium parent, then the 235U 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 235U 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 235U enrichment of fuel rods. This new technique also makes the gamma-ray passive method perfectly. (11 figs., 6 tabs.)

  17. Plate-shaped high power nuclear fuel element containing low enrichment uranium and its preparation

    International Nuclear Information System (INIS)

    The present invention provides a plate-shaped high power nuclear fuel element containing low enrichment uranium (5 to 20 percent by weight uranium235 in the uranium component) as the fissionable material, the fuel element essentially comprising a plate of UAl4 provided with a sheath (clad) of aluminum or an aluminum alloy and impurities inherent to the manufacturing process. (DG)

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

    International Nuclear Information System (INIS)

    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 UAlx 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 UAlx 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 Alx 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-01

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

  20. Qualification in the reactor Siloe of low enriched fuels for research and test reactors

    International Nuclear Information System (INIS)

    For nearly two years now, in the scope of the Reduced Enrichment Research and Test Reactor (RERTR) Program (CEA-ANL-CERCA Agreements), low enriched fuel has been irradiated in Siloe. In 1981 a complete 45% enriched fuel element (U Alx compound) was irradiated. A burn-up of 50% was obtained without any difficulty. Since June 1982 4 U3Si fuel plates are being irradiated. These plates, with a density of 5.5g of total uranium per cubic centimeter (two plates), and 6.0g per cubic centimeter (the other two plates) have already reached a mean burn-up of about 20% and their behaviour up till now is excellent. The fuel element and the plates have been manufactured by CERCA

  1. Feasibility study of using of low enriched uranium fuel for research reactor in Sofia

    International Nuclear Information System (INIS)

    The results of the study of arrangement of the reactor core of pool type Research Reactor 200 kW in Sofia are presented. The number of planned horizontal experimental channels of Research Reactor is equal to 8 and vertical ones to 9. The beginning of study performed for fresh IRT2M tubular fuel assemblies containing highly enriched uranium with 36% enrichment is presented. The core modelling calculations are performed by the MCNP4C code. The IRT-2M fuel assembly modelling is tested by experimental data. The configuration of the reactor core including three-tube and four-tube fuel assemblies as well as beryllium blocks is analysed. The aspects of feasibility of using of developed in Russia pin-type fuel assemblies IRT-MR with low enriched uranium (LEU<20%) in core design instead of IRT-2M ones are discussed. (author)

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

    International Nuclear Information System (INIS)

    Russia possesses considerable experience in utilisation of uranium-molybdenum alloys containing in dispersion fuel composition no more than 6 g/cm3 uranium. The feasibility of utilising the U-9 mass.% Mo alloy with reduced enrichment uranium (< 20%) in research reactor dispersion fuel pins has been analysed in the IPPE. Specimens with the 40 vol.% (U-9 mass. % Mo) + 60 vol.% Al fuel have been fabricated by hot pressing. Investigations of thermal physical properties of this fuel as well as tests for compatibility of U-Mo alloy with Al have been carried out in a wide temperature range. Corrosive tests of dispersion fuel have been realised in water. A flow chart of reproducing wastes from fuel pin production has been considered. The results of works carried out enable to hope on successful solution of the problem of utilisation high-density U-Mo fuel in research reactors. (author)

  3. 76 FR 26996 - Notice of Intent To Prepare an Environmental Impact Statement for New Corporate Average Fuel...

    Science.gov (United States)

    2011-05-10

    ... Environmental Impact Statement for New Corporate Average Fuel Economy Standards AGENCY: National Highway Traffic... consider the potential environmental impacts of new fuel economy standards for model years 2017-2025... and the Administrator of the Environmental Protection Agency (EPA), to establish average fuel...

  4. 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)

  5. A Fast Test for Excessive U235 Concentrations in Enriched Fuel Elements

    International Nuclear Information System (INIS)

    This paper describes a new non-destructive method of detecting areas of enriched fuel elements containing excessive U235 , and an instrument designed to inspect uranium-aluminium alloy fuel for the NRU reactor using the new method. The new method is as precise as conventional methods but is m u c h faster. The improvement in test speed depends on the application, but typically m a y be a factor of four. Alternatively, the method m a y be used to improve the precision of the test without increasing the testing time. The method may be used when the U235 concentration in selected areas of a fuel element is estimated by measuring the intensity of 184-keV γ-rays (from the natural decay of U235 ) and when an area is considered acceptable if the measured intensity from it is less than a limiting intensity equivalent to the maximum allowable concentration. It the measured intensity is much less or much greater than the limiting intensity, much less time is required to make this decision than when they are nearly the same. The new method takes advantage of this. It thus inspects each area only as long as is necessary to make a decision with the required confidence. The average time taken in practical cases is less than that taken by conventional methods which inspect each area for a pre-selected time chosen to give the required confidence in the most difficult cases - when the measured and limiting intensities are nearly the same. The new method is based on sequential sampling theory. It may also be used to detect areas with undesirably low U235 concentration. (author)

  6. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

  10. Study of application and testing of the experimental fuel assembly with 36 %, 19.8% enrichment

    International Nuclear Information System (INIS)

    Study of application and testing of the experimental fuel assembly with 36% enrichment in U-235, the fuel developed by Russian Federation at the WWR-SM research reactor of INP - Institute of Nuclear Physics (the Republic of Uzbekistan). The WWR-SM reactor of INP currently uses FA of the IRT-3M type containing uranium with an enrichment of 36% in U-235. The density of uranium in the fuel meat is 2.5 g/cm3. The reactor has a high operating efficiency (6400 hours per year) at a maximum power of 10 MW. For research reactors in countries of the Former Soviet Union (FSU), nuclear fuel assemblies (FA) containing low-enriched uranium (LEU), which meet the international requirements, have not yet been developed. Scientists at Institute 'Kurchatov' together with scientists at the INP are currently developing the fuel elements with an enrichment of 36%, (19.8% IRT-4M type) in U-235 that meets the requirements of IAEA. (author)

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

    International Nuclear Information System (INIS)

    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.)

  12. Plutonium-enriched thermal fuel production experience in Belgium

    International Nuclear Information System (INIS)

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

  13. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

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

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

    International Nuclear Information System (INIS)

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U3Si in aluminum, to complement the dispersions of U3Si 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% U3Si have been manufactured. (author)

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

    International Nuclear Information System (INIS)

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

  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. Neutron flux measurements in a core formed by low enrichment uranium (LEU), assembly fuel

    International Nuclear Information System (INIS)

    The RECH-1 reactor started operation with highly enriches fuel (80% 235U); nevertheless, due to international agreements, the next charge ought to consider a reduction in the enrichment level, from 45 % to 20 % in 235U, to reach, later on, the accepted international level of 19.75 %. The RECH-1 enrichment conversion process has been gradual; is to say, it went from high to medium through mixed cores and the same strategy is being used to go from medium to low enrichment. With the purpose of making characterization measurements of a core formed only with low enrichment, and so to compare results with equivalent cores using MHU or HEU, it was decided to shape a core entirely conformed with by 32 new low enrichment, made at the CCHEN's Chilean Fuel Fabrication Plant (PEC). Amongst all measurements done, it can be found the thermal, epithermal and fast neutron flux profiles in positions of interest, making use for it Au and In activation probes. The main object of this paper is to show the experimental results obtained, which, just in comparison way, will be contracted to values achieved using MCNP neutron calculations, based in Montecarlo's theory (Au)

  20. Choice and use of a low-level enriched fuel in high performance research reactors

    International Nuclear Information System (INIS)

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

  1. Choice and utilization of a moderately enriched nuclear fuel in high performance research reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Description is made of the processes used in the production of U3O8 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)

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

    International Nuclear Information System (INIS)

    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

  4. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-10-24

    ..., New Mexico, and has authorized the introduction of uranium hexafluoride (UF 6 ) into cascades numbered... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... 4th day of October, 2013. For the U.S. Nuclear Regulatory Commission. Brian W. Smith, Chief,...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bakhtyar, S.; Iqbal, M.; Israr, M.; Pervez, S.; Salahuddin, A. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

  9. Effect of high-density fuel loading on criticality of low enriched uranium fueled material test research reactors

    International Nuclear Information System (INIS)

    The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less 235U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in 235U loading. All this was achieved with

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  14. The contamination of the low enriched uranium fuel with oxygen during the manufacturing process

    International Nuclear Information System (INIS)

    The manufacturing of TRIGA fuel rods with low enriched uranium (≤ 20% 235U) follows in principle the same route as for high-enriched uranium (93% 235U). The contamination with chemical elements during the fabrication process deteriorates the fuel properties and its quality required by the run in the reactor. From the standpoint of the phase relations, the most important impurity is the oxygen. The oxygen concentration influences the kinetics of the zirconium hydriding process. If during the hydriding process contamination with oxygen occurs a decrease of the hydriding rate will take place. This paper presents the aspects of the TRIGA fuel contamination with oxygen during manufacturing process and ways to reduce it. The permanent control of the oxygen concentration in the working zone avoids the accidental contamination. (authors)

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

    International Nuclear Information System (INIS)

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

  16. 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

  17. 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

  18. Neutronics calculations relevant to the conversion of research reactors to low-enriched fuel

    International Nuclear Information System (INIS)

    The new spirit and urgency of converting the remaining research reactors from highly enriched uranium (HEU) to low-enriched fuel, combined with the prospects of new ultra-high-density fuels, provides the main impetus and defines the basic scientific objectives for this thesis. It is predictable that activities to convert existing research reactors will intensify in the near-term future, which in turn would simultaneously increase the need for corresponding neutronics calculations. Here, especially the analysis of the remaining high-flux reactors, which are most difficult to convert due to compact core geometries, may benefit from high-precision simulation tools to adequately set-up and study reactor parameters using complete three-dimensional core models. The scope of the present thesis is to support this process in providing a new computational tool for neutronics calculations (M3O), which is based on standard physics codes, while using the technical computing environment Mathematica as the primary user-interface. The use of such modern environments can be very convenient for a variety of reasons: their analytical capabilities allow for a broad range of calculations and data manipulation, while their interactive graphical user-interface facilitates intensive control of input parameters and interpretation of achieved results. At the same time, Monte Carlo methods play an increasing role in neutron transport and burnup analyses. In M3O, the Monte Carlo code MCNP is employed, which offers the potential for high-precision modeling and analysis. Both major components, Mathematica and MCNP, are also used in an optimization tool developed below and based on the linear programming technique to optimize reactor performance by variation of the fundamental core parameters. The potential (and limits) of monolithic fuels is largely unknown today. Even though the conversion of a large number of medium-flux reactors would be relatively straightforward, the performance of

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 UO2 or ThO2 fertile material were located in the fuel zone of the blanket and helium gas or Flibe (Li2BeF4) 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 UO2, respectively and for operation periods of 18 and 48 months in case of fuel rods containing ThO2, 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

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

    International Nuclear Information System (INIS)

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

  2. Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs

    International Nuclear Information System (INIS)

    Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  7. Functionally stable and phylogenetically diverse microbial enrichments from microbial fuel cells during wastewater treatment.

    Science.gov (United States)

    Ishii, Shun'ichi; Suzuki, Shino; Norden-Krichmar, Trina M; Nealson, Kenneth H; Sekiguchi, Yuji; Gorby, Yuri A; Bretschger, Orianna

    2012-01-01

    Microbial fuel cells (MFCs) are devices that exploit microorganisms as biocatalysts to recover energy from organic matter in the form of electricity. One of the goals of MFC research is to develop the technology for cost-effective wastewater treatment. However, before practical MFC applications are implemented it is important to gain fundamental knowledge about long-term system performance, reproducibility, and the formation and maintenance of functionally-stable microbial communities. Here we report findings from a MFC operated for over 300 days using only primary clarifier effluent collected from a municipal wastewater treatment plant as the microbial resource and substrate. The system was operated in a repeat-batch mode, where the reactor solution was replaced once every two weeks with new primary effluent that consisted of different microbial and chemical compositions with every batch exchange. The turbidity of the primary clarifier effluent solution notably decreased, and 97% of biological oxygen demand (BOD) was removed after an 8-13 day residence time for each batch cycle. On average, the limiting current density was 1000 mA/m(2), the maximum power density was 13 mW/m(2), and coulombic efficiency was 25%. Interestingly, the electrochemical performance and BOD removal rates were very reproducible throughout MFC operation regardless of the sample variability associated with each wastewater exchange. While MFC performance was very reproducible, the phylogenetic analyses of anode-associated electricity-generating biofilms showed that the microbial populations temporally fluctuated and maintained a high biodiversity throughout the year-long experiment. These results suggest that MFC communities are both self-selecting and self-optimizing, thereby able to develop and maintain functional stability regardless of fluctuations in carbon source(s) and regular introduction of microbial competitors. These results contribute significantly toward the practical application

  8. Functionally stable and phylogenetically diverse microbial enrichments from microbial fuel cells during wastewater treatment.

    Directory of Open Access Journals (Sweden)

    Shun'ichi Ishii

    Full Text Available Microbial fuel cells (MFCs are devices that exploit microorganisms as biocatalysts to recover energy from organic matter in the form of electricity. One of the goals of MFC research is to develop the technology for cost-effective wastewater treatment. However, before practical MFC applications are implemented it is important to gain fundamental knowledge about long-term system performance, reproducibility, and the formation and maintenance of functionally-stable microbial communities. Here we report findings from a MFC operated for over 300 days using only primary clarifier effluent collected from a municipal wastewater treatment plant as the microbial resource and substrate. The system was operated in a repeat-batch mode, where the reactor solution was replaced once every two weeks with new primary effluent that consisted of different microbial and chemical compositions with every batch exchange. The turbidity of the primary clarifier effluent solution notably decreased, and 97% of biological oxygen demand (BOD was removed after an 8-13 day residence time for each batch cycle. On average, the limiting current density was 1000 mA/m(2, the maximum power density was 13 mW/m(2, and coulombic efficiency was 25%. Interestingly, the electrochemical performance and BOD removal rates were very reproducible throughout MFC operation regardless of the sample variability associated with each wastewater exchange. While MFC performance was very reproducible, the phylogenetic analyses of anode-associated electricity-generating biofilms showed that the microbial populations temporally fluctuated and maintained a high biodiversity throughout the year-long experiment. These results suggest that MFC communities are both self-selecting and self-optimizing, thereby able to develop and maintain functional stability regardless of fluctuations in carbon source(s and regular introduction of microbial competitors. These results contribute significantly toward the

  9. Neutronic calculations of PARR-1 cores using leu-silicide fuel. [leu (low enriched uranium)

    Energy Technology Data Exchange (ETDEWEB)

    Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

    1991-08-01

    Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing Low Enriched Uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full power operation and the equilibrium cores. The burnup study of the equilibrium core and calculations for discharged fuel inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis.

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

    International Nuclear Information System (INIS)

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

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  15. RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

    Energy Technology Data Exchange (ETDEWEB)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  18. Experimental operation of the RA reactor with 4 fuel channels containing 80% enriched dispersion fuel - Operational Report

    International Nuclear Information System (INIS)

    Start of utilization of the new 80% enriched dispersion nuclear fuel is underway in the RA reactor core. Both economic and technical analyses were in favor of introducing the new fuel elements gradually into the RA reactor core. Thus overall theoretical and experimental analyses as well as other preparations are directed to transition regime based on gradual introducing of new fuel into the core, i.e. reactor core with two types of fuel. The objective of these analyses and preparation is establishment of conditions for safe reactor operation during transition period. The analyses and preparations are almost completed. The experimental data about fuel burnup during a time period of operation at nominal power i.e. daily decrease of excess reactivity is missing. This data is needed for planning the refueling (quantity of fresh fuel and frequency of refueling) during the transient period. This data can be obtained only by normal operation of the reactor during a period of time significantly longer than the period of attaining equilibrium poisoning, as time between two D2O condensate overflows into the RA reactor core. Thus a ten day experimental campaign was planned to be done in December 1976. This report presents the most important results of safety analyses and preparation which show that, during this experimental period, the reactor operation is absolutely safe taking into account the most important parameters influencing reactor safety, as reactivity, thermal and temperature limits for fuel and the reactor, etc. Data to be obtained during this experimental campaign are significant because they would enable definition of future supply of fresh fuel during the transition period

  19. Atomics International fuel fabrication facility and low enrichment program. Part 1

    International Nuclear Information System (INIS)

    The Al facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAlx 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 UAlx powder line into operation and we have had to expand some of our storage area

  20. 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.

  1. 77 FR 62623 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Science.gov (United States)

    2012-10-15

    ... Fuel Economy Standards AGENCIES: Environmental Protection Agency (EPA) and National Highway Traffic... October 15, 2012 Part II Environmental Protection Agency 40 CFR Parts 85, 86, and 600 Department of... Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy...

  2. Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

    International Nuclear Information System (INIS)

    Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm3 by the fuel of density 4.00 gU/cm3 then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm3 in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29-62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1

  3. Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad 45650 (Pakistan); Aslam [Department of Physics and Applied Mathematics, PIEAS, Islamabad 45650 (Pakistan)]. E-mail: aslam_mcmaster@yahoo.com; Ahmad, Nasir [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad 45650 (Pakistan)

    2005-07-15

    Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm{sup 3} by the fuel of density 4.00 gU/cm{sup 3} then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm{sup 3} in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29-62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  9. 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.

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

    International Nuclear Information System (INIS)

    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/cm3, 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)

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

    International Nuclear Information System (INIS)

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

  12. Pakistan upgrades PARR-1 and converts to LEU. [Upgrading of research reactor and conversion to low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-12-01

    The Pakistan Research Reactor, PARR-1, is a 5MW swimming pool type reactor originally designed to use MTR type fuel elements fabricated from uranium enriched to more than 90%. After about 24 years of satisfactory operation it is now planned to convert the reactor to use low enriched (20%) uranium fuel. The opportunity will also be taken to upgrade the reactor power to about 9MW. This power upgrading will meet the demand for higher neutron fluxes for experimental and radioisotope production as well as compensating for the neutron flux penalty arising from conversion from high enriched to low enriched fuel. During the process of conversion and upgrading it is also proposed to renovate existing services and associated systems and to add certain new safety related engineering. (author).

  13. Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

    International Nuclear Information System (INIS)

    Highlights: • A simulation model for KUR LEU silicide core was established. • Safety analyses for reactivity insertion transients were performed by EUREKA-2/RR. • Accidental control rod withdrawal transients were analyzed. • Cold water injection induced reactivity insertion transients were analyzed. • Reactivity insertion transients due to removal of irradiation samples were analyzed. - Abstract: The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2 gU/cm3) and enrichment (20%) of LEU fuel (U3Si2–AL) are quite different from the uranium density (0.58 gU/ (cm3)) and enrichment (93%) of HEU fuel (U–Al), which may result in the changes of heat transfer response and neutronic characteristic in the core. So it is necessary to objectively re-assess the feasibility of LEU silicide fuel core in KUR by using various numerical simulation codes. This paper established a detailed simulation model for the LEU silicide core and provided the safety analyses for the reactivity insertion transients in the core by using EUREKA-2/RR code. Although the EUREKA-2/RR code is a proven and trusted code, its validity was further confirmed by the comparison with the predictions from another two thermal hydraulic codes, COOLOD-N2 and THYDE-W at steady state operation. The steady state simulation also verified the feasibility of KUR to be operated at rated thermal power of 5 MW. In view of the core loading patterns, the operational conditions and characteristics of the reactor protection system in KUR, the accidental control rod withdrawal transients at natural circulation and forced circulation modes, the cold water injection induced reactivity insertion transient and the

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

  19. 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.

  20. In core fuel management optimization by varying the equilibrium cycle average flux shape for batch refuelled reactors

    International Nuclear Information System (INIS)

    We suggest a method to overcome this problem of optimization by varying reloading patterns by characterizing each particular reloading pattern by a set of intermediate parameters that are numbers. Plots of the objective function versus the intermediate parameters can be made. When the intermediate parameters represent the reloading patterns in a unique way, the optimum of the objective function can be found by interpolation within such plots and we can find the optimal reloading pattern in terms of intermediate parameters. These have to be transformed backwards to find an optimal reloading pattern. The intermediate parameters are closely related to the time averaged neutron flux shape in the core during an equilibrium cycle. This flux shape is characterized by a set of ratios of the space averaged fluxes in the fuel zones and the space averaged flux in the zone with the fresh fuel elements. An advantage of this choice of intermediate parameters is that it permits analytical calculation of equilibrium cycle fuel densities in the fuel zones for any applied reloading patten characterized by a set of equilibrium cycle average flux ratios and thus, provides analytical calculations of fuel management objective functions. The method is checked for the burnup of one fissile nuclide in a reactor core with the geometry of the PWR at Borssele. For simplicity, neither the conversion of fuel, nor the buildup of fission products were taken into account in this study. Since these phenomena can also be described by the equilibrium cycle average flux ratios, it is likely that this method can be extended to a more realistic method for global in core fuel management optimization. (orig./GL)

  1. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels. Guidebook addendum: Heavy water moderated reactors

    International Nuclear Information System (INIS)

    A Guidebook on Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels (IAEA-TECDOC--233) was issued by the International Atomic Energy Agency in August 1980. This document contains a wide variety of information of the physics, thermal-hydraulics, fuels, and fuel cycle economics for light water moderated research and test reactors. In consideration of the special features of heavy water moderated research and test reactors (hereafter referred to as heavy water research reactors), this Addendum to IAEA-TECDOC--233 has been prepared to assist operators and physicists from these reactors in determining whether conversion from HEU to 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. The organization of this Addendum follows that of IAEA-TECDOC--233 as closely as possible in order to provide a consistent presentation of the information and to minimize the repetition of information that is common to both heavy water and light water research reactors. Distinctive features of the heavy water reactors are addressed where applicable

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

    International Nuclear Information System (INIS)

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

  3. Influence of N-15 enrichment on neutronics, costs and C-14 production in nitride fuel cycle scenarios

    International Nuclear Information System (INIS)

    The C-14 production for different closed fuel cycle scenarios has been investigated. If nitride fuel is used in fast reactors and ADSs dedicated to management of plutonium and minor actinides, an N-15 enrichment level of about 99% is required for the nitride cores to produce the same amount of C-14 as the oxide cores in the power park. The corresponding cost penalty for fuel fabrication is estimated to be larger than 25%. If reprocessing is included in the costs for fuel operations, the penalty is of the order of 5-10%, provided that a closed gas cycle is implemented for the fabrication. If nitride fuels are used only for minor actinide management in ADS, the required enrichment level is about 93%, and the cost penalty is less than 10%. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 (UO2) in aluminum master form and the UA14 alloy. Preliminary calculations have shown that, with the WWR-K reactor core preserved existing critical characteristics (the fuel composition: UA14), 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

  6. Plutonium credit in the low enriched uranium fuel of PARR-1

    Energy Technology Data Exchange (ETDEWEB)

    Tayyab, M. [Pakistan Institute of Nuclear Science and Technology, Nuclear Engineering Division, P. O. Nilore, Islamabad (Pakistan)]. E-mail: mtmalikpk@yahoo.com; Bakhtyar, S. [Pakistan Institute of Nuclear Science and Technology, Nuclear Engineering Division, P. O. Nilore, Islamabad (Pakistan); Hamid, T. [Pakistan Institute of Engineering and Applied Sciences, P. O. Nilore, Islamabad (Pakistan)

    2004-08-01

    The amount of plutonium (Pu) isotopes and the resultant savings of {sup 235}U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations. The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 x 10{sup -3}{delta}k/k. Amount of {sup 235}U equivalent to this value of reactivity was found to be 15.58 {+-} 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box.

  7. Plutonium credit in the low enriched uranium fuel of PARR-1

    International Nuclear Information System (INIS)

    The amount of plutonium (Pu) isotopes and the resultant savings of 235U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations. The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 x 10-3Δk/k. Amount of 235U equivalent to this value of reactivity was found to be 15.58 ± 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box

  8. ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; National Security; Inst. of Physics and Power Engineering

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation

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

    International Nuclear Information System (INIS)

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

  10. The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

    International Nuclear Information System (INIS)

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  15. 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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-26

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

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

    KAUST Repository

    Crumlin, Ethan J.

    2012-01-01

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

  20. 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.

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

    International Nuclear Information System (INIS)

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

  2. Bulgarian Experience with the Implementation of 235U Higher Enriched Fuel in WWER-1000 (Water Chemistry Aspects)

    International Nuclear Information System (INIS)

    Water chemistry and radiochemistry plant data from the WWER-1000 Units in NPP Kozloduy confirm that a realistic way for satisfactory implementation of 235U high enriched (up to 4.3%) fuel has been found. The main requirements are: implementation of solid neutron burnable absorbers; application of corrosion resistant fuel cladding; and the maintenance of suitable coolant water chemistry. The implementation at NPP Kozloduy is described. (author)

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

    International Nuclear Information System (INIS)

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

  4. Modeling of a highly enriched 235U fission chamber for spent fuel assay

    International Nuclear Information System (INIS)

    Highlights: • Accurate fission chamber models require accurate design information. • Fissile mass amount, active layer thickness and structural materials determine the detector sensitivity. • Fission products in fission chambers can be modeled and transported in a Monte Carlo code such as MCNPX. • Fission products transport plays a key role when determining a fission chamber sensitivity. • The model is in good agreement with experimental data. - Abstract: Fission chambers loaded with high enriched uranium are used for spent fuel measurements in the so-called Fork detector. The Fork detector is one of the work-horses used by safeguards inspectors for spent fuel measurements during verification activities in the framework of the Non-Proliferation and Euratom Treaties. Having an accurate and validated model of the measurement equipment is beneficial for the investigation of this type of applications. SCK• CEN is carrying out a significant effort to model the Fork detector with the MCNPX code. However, scarce information is known about the fission chambers. This work describes the impact of the design information of the fission chamber on the calculated detector sensitivity and, consequently, on the overall Fork detector response for neutrons, using Monte Carlo simulations. The heavy ions transport in the active layer of the fission chamber was also studied and the resulting fission product energy spectra were compared with the available experimental data

  5. Reactor physics studies leading to a fuel cost survey for an HTR system with low U235 enrichment

    International Nuclear Information System (INIS)

    Reactor physics studies have been carried out on an HTR system with low U235 enrichment. The work reported establishes the total fuel cost as approximately Pound28/kW and provides sufficient information for an overall plant optimisation. (author)

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

    International Nuclear Information System (INIS)

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

  7. Comparison of some measurements in the Greek Research Reactor-1 (GRR-1) with 20% and 93% enriched uranium fuel elements

    International Nuclear Information System (INIS)

    Experimental results of the core performance of the GRR-1 with 20% and 93% enriched 235U are compared. In particular, the rod worth, the flux distribution (axial and radial), the temperature, and the void effects are analyzed and compared for the two cores. The experimental results are in good agreement with benchmark calculations for this reactor and for other similar reactors. Although the fuel elements of the two cores do not have the same number of plates, the results are of special significance, since they present the only experimental comparative work known so far on the core conversion problem from the use of highly enriched uranium to the use of low enriched uranium fuels

  8. Enrichment of Microbial Electrolysis Cell Biocathodes from Sediment Microbial Fuel Cell Bioanodes

    Energy Technology Data Exchange (ETDEWEB)

    Pisciotta, JM; Zaybak, Z; Call, DF; Nam, JY; Logan, BE

    2012-07-18

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

  9. Program for qualifying CNEA (National Atomic Energy Commission) as a manufacturer of low enriched uranium silicide fuels

    International Nuclear Information System (INIS)

    Full text: This report presents the program for the production and irradiation of a low enriched uranium (LEU) fuel element containing a dispersion of U3Si2 particles in an Al matrix, with a total uranium content of 4.8 g/cm3. The project is being carried out by the Nuclear Fuels Department of the National Atomic Energy Commission (CNEA) of Argentina and aims at qualifying our organization as manufacturer of LEU fuel elements for research reactors. The program involves the design, fissile material production, components fabrication and inspection, assembly, irradiation and postirradiation tests of two prototypes. The meat will be a dispersion of U3Si2 in an Al matrix with a total uranium content of 4.8 g/cm3. The irradiations will be performed at the RA-3. The first prototype is conceived in such a way to facilitate the posterior disassembly and PIE examinations. The second one will have the design of the future normal fuel elements of the RA-3. This project relies on the experience in CNEA in the production of standard fuel elements for the RA-3 and of uranium silicide mini plates successfully tested at the Oak Ridge Research Reactor within the RERTR program (Reduced Enrichment for Research and Test Reactors). Important investments have already been made in the installations for fuel elements production and PIE

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

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Phuong Tran

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Soyama, Kazuhiko; Ichikawa, Hiroki; Kodaira, Tsuneo (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1992-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 UPu0.45O2, UPu0.45N, UPu0.6N, PuN + ZrN, PuO2 + 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

  20. 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

  1. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  2. Good practices for qualification of high density low enriched uranium research reactor fuels

    International Nuclear Information System (INIS)

    This good practices publication has been prepared to provide points of reference for the type, quality, and completeness of information to be generated in order to ensure the acceptable performance of high density LEU fuels to be used in existing and new training, research, test, and isotope production reactors. The IAEA anticipates that the information presented here will be of value to fuel developers, reactor operators planning to use a new fuel, and to regulatory bodies faced with deciding whether a specific reactor can be licensed to use a new fuel. This good practices publication addresses basic definitions, approaches, and processes relevant to the qualification of research reactor fuel, and defines essential information required for the licensing and use of fuels in research reactors. It concentrates on development and qualification of high density fuels of the type used in most research and test reactors, i.e., fuel consisting of a fuel meat contained within a metallic cladding. Any other type of fuel, e.g. a homogeneous solution fuel, is outside the scope of this publication. Essential definitions are presented in Section 3. In Section 4, a description of the type and extent of information needed to support a fuel qualification report is presented, while many additional pointers are given in Appendix I. An overview of the qualification process is presented in Section 5. The similar and somewhat flexible qualification processes employed for research reactor fuels in Argentina, Canada, France, the Republic of Korea, and the United States of America on one hand and the very structured and codified process used in the Russian Federation on the other hand are described in Section 6. Examples and case histories for both types of approaches are presented in Appendix II. A list of the acronyms used in this document is presented in Appendix III

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

    International Nuclear Information System (INIS)

    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 UO2 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; UO2 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. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    International Nuclear Information System (INIS)

    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 (Pw, =5, 6, 7, 8, 9, and 10 MWm-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 (Li2BeF4), natural lithium (Li), and eutectic lithium (Li17Pb83). Investigations are observed during 4 years for discrete time intervals ofΔt= 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)

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

    Science.gov (United States)

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

    2016-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  8. 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.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Calculated research for conversion of research reactor in Uzbekistan to low enriched fuel

    International Nuclear Information System (INIS)

    Full text: The paper comprises the results of calculations for conversion of nuclear reactor VVR-CM (located in Uzbekistan) from HEU to LEU by means of tube and rod type fuel elements with U-Mo alloy nuclear fuel. (author)

  15. Uranium enrichment

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Improving Power Production in Acetate-Fed Microbial Fuel Cells via Enrichment of Exoelectrogenic Organisms in Flow-Through Systems

    International Nuclear Information System (INIS)

    An exoelectrogenic, biofilm-forming microbial consortium was enriched in an acetate-fed microbial fuel cell (MFC) using a flow-through anode coupled to an air-cathode. Multiple parameters known to improve MFC performance were integrated in one design including electrode spacing, specific electrode surface area, flow-through design, minimization of dead volume within anode chamber, and control of external resistance. In addition, continuous feeding of carbon source was employed and the MFC was operated at intermittent high flows to enable removal of non-biofilm forming organisms over a period of six months. The consortium enriched using the modified design and operating conditions resulted in a power density of 345 W m-3 of net anode volume (3650 mW m-2), when coupled to a ferricyanide cathode. The enriched consortium included -, -, -Proteobacteria, Bacteroidetes and Firmicutes. Members of the order Rhodocyclaceae and Burkholderiaceae (Azospira spp. (49%), Acidovorax spp. (11%) and Comamonas spp. (7%)), dominated the microbial consortium. Denaturing gradient gel electrophoresis (DGGE) analysis based on primers selective for Archaea suggested a very low abundance of methanogens. Limiting the delivery of the carbon source via continuous feeding corresponding to the maximum cathodic oxidation rates permitted in the flow-through, air-cathode MFC resulted in coulombic efficiencies reaching 88 5.7%.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Adams, T.M.

    1998-11-25

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

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

    Science.gov (United States)

    Vamshi Krishna, K; Venkata Mohan, S

    2016-08-01

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

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

    International Nuclear Information System (INIS)

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

  2. High density LEU [low enriched uranium] fuel development at Babcock and Wilcox

    International Nuclear Information System (INIS)

    An aggressive pursuit of developing a high-density LEU fuel process has been undertaken over the past six years at the Babcock and Wilcox Co. A major effort has been devoted to the U3Si2 fuel development. Today B and W feels confident that their current U3Si2 manufacturing process is comparable to existing U3O8 and UAlx fuel technologies. A continued effort will be maintained within the U3Si2 product line to provide the highest product quality and to increased process efficiencies. Investigations into other high density LEU fuel development such as U(x)Si(y) alloys will only be secondary considerations. (Author)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    G. Youinou; S. Bays

    2009-05-01

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

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

    International Nuclear Information System (INIS)

    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)

  6. Enhancing clostridial acetone-butanol-ethanol (ABE) production and improving fuel properties of ABE-enriched biodiesel by extractive fermentation with biodiesel.

    Science.gov (United States)

    Li, Qing; Cai, Hao; Hao, Bo; Zhang, Congling; Yu, Ziniu; Zhou, Shengde; Chenjuan, Liu

    2010-12-01

    The extractive acetone-butanol-ethanol (ABE) fermentations of Clostridium acetobutylicum were evaluated using biodiesel as the in situ extractant. The biodiesel preferentially extracted butanol, minimized product inhibition, and increased production of butanol (from 11.6 to 16.5 g L⁻¹) and total solvents (from 20.0 to 29.9 g L⁻¹) by 42% and 50%, respectively. The fuel properties of the ABE-enriched biodiesel obtained from the extractive fermentations were analyzed. The key quality indicators of diesel fuel, such as the cetane number (increased from 48 to 54) and the cold filter plugging point (decreased from 5.8 to 0.2 °C), were significantly improved for the ABE-enriched biodiesel. Thus, the application of biodiesel as the extractant for ABE fermentation would increase ABE production, bypass the energy intensive butanol recovery process, and result in an ABE-enriched biodiesel with improved fuel properties. PMID:20585897

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

    International Nuclear Information System (INIS)

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

  8. Evaluation of the feasibility of applying cermet fuel pins on basis of uranium enriched UP to 20% 235U to up-rated power research reactors

    International Nuclear Information System (INIS)

    The paper deals with the feasibility of employing fuel pins with a cermet fuel composition (UO2 particulates in an Al-based alloy) clad with the zirconium-based Zr+1%Nb alloy to elevated power research reactors. The fuel pin production process enables to place uranium dioxide particulates, up to 65% by volume, into small diameter claddings and to attain in fuel composition the volume densities up to 6.4 g U/cm3 of uranium. The impregnation of the charged UO2 particulates with the liquid matrix material in the subsequent course of fuel pin fabrication enables to ensure high thermal conductivity of the cermet fuel composition and zero thermal resistance at the fuel-cladding interface. The computations (made, by way of example, with reference to the IR-8 research reactor) have shown that the replacement of tubular fuel elements with high-enriched uranium, by fuel pins whose cermet fuel is not enriched over 20% is possible in principle, with the main thermal-hydraulics and neutronics characteristics of the reactor being conserved. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-13

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 235U nuclide per TVR-S FE. (author)

  13. Progress in postirradiation examination and analysis of low-enriched U-Mo research reactor fuels

    International Nuclear Information System (INIS)

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL Alpha-Gamma Hot Cell Facility. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm-3 atomized and machined uranium alloy powders containing 6.5 wt% to 10 wt% molybdenum. In addition, two miniplates containing solid U-10wt% Mo foils. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from a companion test performed to 50% burnup, RERTR-5, are presented. (author)

  14. Failure of a TRIGA low enriched uranium (LEU) fuel during long term burnup testing in the Oak Ridge Reactor (ORR)

    International Nuclear Information System (INIS)

    The development of higher loaded LEU uranium zirconium-hydride fuel culminated in long term burn-up testing in the Oak Ridge Research Reactor (ORR). A sixteen rod cluster was designed as the test article containing ER-U-ZrH1.6 fuel pins clad with Incoloy 800. The nominal fuel pin geometry is 0.542 in. (13.77 mm) outside diameter, 22.0 in (559 mm) active length, with a cladding thickness of 0.016 in. (0.406 mm). Irradiation of the TRIGA-LEU cluster in the ORR began on 13 December 1979 and is still in progress. A total of 19 test rods have been included in the irradiation campaign; nine with 45 wt-% uranium (3.7 gms/cc), six with 30 wt-% and four with 20 wt-% - all are enriched to 19.7 percent in uranium 235. The irradiation of the 20 and 30 wt-% rods was terminated in May 1982 after the targeted burn-up levels of 35 and 40% respectively were successfully attained. One of the 45 wt-% rods (not the highest burnup rod) failed in November 1983 as a result of internal pressure causing the clad to split in a typical ductile failure. Of the two possible modes of failure - excess steam pressure or excess hydrogen pressure - steam pressure is the only really probable means of failure. Fuel temperatures would have had to exceed 1150 deg C to produce the necessary hydrogen pressure to cause failure and temperatures that high are extremely improbable without film blanketing at the clad surface. There was no evidence of film blanketing which would have caused obvious clad discoloration. The best explanation for the failure is that a very small clad leak developed in the region where the major failure subsequently occurred. The clad leak was likely the result of a clad imperfection manifesting itself after four years of irradiation. Fission gas release from the small leak resulted in a decision to shut down the ORR. The test fuel rod became water-logged during this shutdown period and during the subsequent startup of the reactor, steam pressure built up. As the power was

  15. Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

  17. Safe use of the Institute of Nuclear Physics reactor with low enriched fuel

    International Nuclear Information System (INIS)

    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)

  18. Flame synthesis of carbon nanostructures using mixed fuel in oxygen-enriched environment

    International Nuclear Information System (INIS)

    The effects of key parameters, namely oxygen concentration, mixed fuel, and sampling positions, on the formation of carbon nano-onions (CNOs) and carbon nanotubes (CNTs) were investigated in oxy-fuel inverse diffusion flames. Particular focus was put on the intermediate species in connection with the synthesis of CNOs and CNTs. Three patterns of carbon nanostructures were observed: CNTs only, CNOs only, and CNTs/CNOs cogeneration. An appropriate temperature range in the synthesis of CNTs was identified to lie between 400 and 1,000 °C, whereas the temperature range for the synthesis of CNOs was higher, within 800–1,250 °C. A threshold value of oxygen concentration, 30 %, existed for onset of CNO synthesis. Gas composition analysis indicated that no carbon nanomaterial was formed at low CO and C2H2 concentration as well as low substrate temperature (lower than 400 °C). Compared with the synthesis condition of CNTs only, the C2H2 concentration was higher for the onset of CNTs/CNOs cogeneration, whereas the CO concentration was maintained at the same level. Additionally, the critical C2H2 concentration for the onset of CNOs only was found to be 0.4 %. A large quantity of CNOs was observed for C2H2 concentration greater than 0.4 % and CO concentration greater than 4 %.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

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

  2. Revisiting the carbonyl n → π* electronic excitation through topological eyes: expanding, enriching and enhancing the chemical language using electron number distribution functions and domain averaged Fermi holes.

    Science.gov (United States)

    Ferro-Costas, David; Francisco, Evelio; Pendás, Ángel Martín; Mosquera, Ricardo A

    2015-10-21

    The theory of chemical bonding is underdeveloped in electronic excited states, even in small molecules. Fortunately, real space tools may be used to offer rich images of simple excitation processes, as is shown in this work. The statistics of electron populations, through a fruitful combination of electron distribution functions (EDFs) and domain averaged Fermi holes (DAFHs), was used to enlighten our chemical knowledge of a paradigmatic process: the n → π* excitation in formaldehyde. Interestingly, our results are perfectly compatible with an alternative perception of the electronic transition: the rotation of one averaged-electron in the oxygen lone pair. This topological model does not require inter-orbital jumps to explain the final electron distribution and, in our humble opinion, this fact makes it, to some extent, more realistic. Finally, other far-reaching conclusions emerge smoothly from our analysis: (i) the σ link may contribute less to the total bond order (as measured by the delocalization index) of a polar double bond than the π one; (ii) populating an antibonding orbital does not necessarily imply decreasing the bond order of its corresponding bond. PMID:26168082

  3. 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)

  4. 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)

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

    International Nuclear Information System (INIS)

    Low-enriched uranium silicide-aluminum dispersion plate-type fuels have been extensively researched and developed under the international program, Reduced Enrichment in Research and Test Reactors. The international effort was led by Argonne National Laboratory (ANL) in the United States. This evaluation is based primarily on reports issued by ANL that discuss and summarize the developmental tests and experiments, including postirradiation examinations, of both miniature and full-sized plates of prototypical fuel compositions. This evaluation concludes that plate-type fuels suitable and acceptable for use in research and test reactors can be fabricated with U3Si2-Al dispersion compacts with uranium densities up to 4.8 g/cm3. 4 refs., 1 fig

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

    Energy Technology Data Exchange (ETDEWEB)

    1988-07-01

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

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

    KAUST Repository

    Abdul Jameel, Abdul Gani

    2016-04-22

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

  8. Uranium Enrichment, an overview

    International Nuclear Information System (INIS)

    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 UF6 for the moment. (author). tabs., figs., 4 refs

  9. Effect of coolant channel width on group constants and multiplication factor of research reactors using MTR type low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, A.N. E-mail: nasir@pieas.edu.pk

    2002-03-01

    One dimensional transport theory lattice code WIMS-D/4 and three dimensional diffusion theory code CITATION have been used to study the effect of fuel loading on critical cores of low enriched uranium (LEU) fuelled material testing reactors (MTRs). The fuel loading in a fuel element was varied by changing the fuel density in the fuel meat. In order to keep the reactor critically moderated, the optimal coolant channel width for a given fuel loading was calculated. For the purpose of optimization, the group constants D, {sigma}{sub a} and {nu}{sigma}{sub f}, and infinite multiplication factor (k{sub {infinity}}) were calculated as a function of coolant channel width using WIMS-D/4. An increase in {sup 235}U loading per fuel plate results in an increase in the optimal coolant channel width and k{sub {infinity}}. The calculated values were found to be in good agreement with the typical design of MTR. CITATION was then used to determine the critical cores for different fuel loading with optimized fuel dimensions. Both critical mass and volume were found to decrease with an increase in the fuel loading. The criticality studies of Pakistan research reactor-1 (PARR-1) are in good agreement with the predictions.

  10. Effect of coolant channel width on group constants and multiplication factor of research reactors using MTR type low enriched uranium fuel

    International Nuclear Information System (INIS)

    One dimensional transport theory lattice code WIMS-D/4 and three dimensional diffusion theory code CITATION have been used to study the effect of fuel loading on critical cores of low enriched uranium (LEU) fuelled material testing reactors (MTRs). The fuel loading in a fuel element was varied by changing the fuel density in the fuel meat. In order to keep the reactor critically moderated, the optimal coolant channel width for a given fuel loading was calculated. For the purpose of optimization, the group constants D, Σa and νΣf, and infinite multiplication factor (k∞) were calculated as a function of coolant channel width using WIMS-D/4. An increase in 235U loading per fuel plate results in an increase in the optimal coolant channel width and k∞. The calculated values were found to be in good agreement with the typical design of MTR. CITATION was then used to determine the critical cores for different fuel loading with optimized fuel dimensions. Both critical mass and volume were found to decrease with an increase in the fuel loading. The criticality studies of Pakistan research reactor-1 (PARR-1) are in good agreement with the predictions

  11. Reliable Fuel Supply and Service Regime Approaches: Restricting the Spread of Enrichment and Reprocessing Through Voluntary Reform of Global Market Structure

    International Nuclear Information System (INIS)

    To enable the expansion of nuclear energy for peaceful purposes while discouraging the spread of enrichment and reprocessing technology to additional countries, existing front and back end supplier States may consider approaches to encourage the establishment of Reliable Fuel Service and Supply (RFS and S) arrangements for providing fresh fuel and taking back spent fuel. Important aspects of such a regime are the economic basis, the product offerings, and alternative business models for RFS and S arrangements. The paper also discusses approaches that might help offset international and domestic political factors that might trump economic considerations when customer States make fuel cycle infrastructure decisions. Specifically addressed in the paper are possible complications from asymmetric business risks posed by the front and back end, impact of current customer State market structure, simple cost estimates of alternative supply and service options, related estimates of incentives to encourage regime development, and impact on incentives of possible future escalations in uranium fuel price. (authors)

  12. 75 FR 80430 - Passenger Car and Light Truck Average Fuel Economy Standards Request for Product Plan Information...

    Science.gov (United States)

    2010-12-22

    ...--classification such as NiMH = Nickel Metal Hydride; Li-ion = Lithium Ion; Li-Air = Lithium Air. b. Battery 100... complete Privacy Act Statement in the Federal Register published on April 11, 2000 (65 FR 19477-78) or you... Intent, 75 FR 62739 (Oct. 13, 2010). \\5\\ Available at http://www.nhtsa.gov/fuel-economy (last...

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

    International Nuclear Information System (INIS)

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

  14. Thorium - denatured uranium fuel cycles in PHWR-pressure tube type using low enriched uranium as annual externally supplied fissile material

    International Nuclear Information System (INIS)

    The use of denatured uranium as the initial fissile inventory of the thorium-uranium cycles could be straight-forward. The use of denatured uranium as annual externally supplied fissile material could be not applicable in the case of homogenous HWR fuel bundle concept if it is intended to avoid the shift from the Th/U cycle to U/Pu cycle or the reenrichment of the recycled uranium containing U-232. The paper presents a heterogenous fuel concept for HWR which permits the use of denatured uranium without the above-mentioned shift. According to this concept the annual externally supplied fissile material is introduced in distinct fuel rods separable at the front end of the reprocessing or as distinct fuel bundles. In these cases the normal reenrichment could be applied, this part of the fuel being free of U-233, and therefore free of U-232. The resource utilization penalties in addition to those introduced by the denaturing of the initial core were evaluated. At 3% enrichment these penalties rise with about 40% the annual natural uranium requirements. It is concluded that for these Th/U cycles in HWR, it is possible to avoid the presence of highly enriched uranium at the fuel fabrication step

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2004-04-01

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

  17. Improved WWER fuel design and manufacturing process, operational experience

    International Nuclear Information System (INIS)

    The WWER fuel design and manufacturing processes applied by OAO MSZ are continuously improved. There are no remarks on the operation of the fuel assemblies with introduced innovations. All the improvements have been justified and are directed at a higher operational reliability and technological effectiveness of the fuel. The transition is in progress to the fuel cycles based on the profiled fuel of 3,82% average enrichment; the profiled fuel assemblies with U/Gd of 4,38% average enrichment are introduced and the second generation fuel of 4,25% average enrichment to be used for a 5-year fuel cycle, as well as the fuel assemblies of 3,82% average enrichment for the scram system with hafnium plates in the joint unit. Besides the above mentioned fuel the profiled U/Gd fuel of 3,84% average enrichment is launched for operation of the scram system. There is a continuous rate of growth of fuel operational time. 71% of the MSZ's fuel assemblies unloaded from the 14 WWER-440 reactors of the second generation (V-213 type) operated during 4 campaigns and more. In 2006 maximum burnup of the WWER-440 fuel was achieved at the Dukovany-4 NPP (49,23 MWd/kgU) for the 3,82% enriched fuel (average). Maximum burnup achieved by the WWER-1000 fuel is 62,24 MWd/kgU (for TVSA operated for 7 campaigns). The average leakage ratio of OAO MSZ's fuel manufactured for the WWER-440 reactors (V-213 type) located outside of Russia and Ukraine over the last 5 years makes 6.6x10-7 and is comparable with the best world achievements. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  19. 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.

  20. 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.

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

    Directory of Open Access Journals (Sweden)

    N. Gruber

    2011-12-01

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

  2. State Averages

    Data.gov (United States)

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

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

    Directory of Open Access Journals (Sweden)

    N. Gruber

    2011-05-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

  7. Fundamental study on serious accidents and their management in fuel fabrication/enrichment facilities and reprocessing facilities

    International Nuclear Information System (INIS)

    The 'Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors' was amended and issued in June 2012 taking into account the lessons derived from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant occurred in March 2011. The main amendments were as follows; Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facility (back-fitting). Japan Nuclear Energy Safety organization (JNES) started this fundamental study on serious accidents and their management, as a safety studying in fuel fabrication/enrichment facilities and reprocessing facilities, for the purpose to contribute to the implementation of new Rules by Nuclear Regulation Authority. From the technical view to be concerned such as fundamental concept of the Rules and applicability of risk-informed regulation, the following 7 subjects were studied: 1) Application concept of the defense in depth to these facilities. 2) Positioning of serious accidents and their management in the defense in depth. 3) Definition of the serious accidents in these facilities. 4) Postulated external events for the study of the serious accidents and their management. 5) Objectives and requirements of the accident management (assurance of reliability). 6) Confirmation logic flow on sequence of the serious accidents and the accident management measures. 7) Applicability of risk information. During the study on these subjects, features of the facilities were clarified at first. Based on concept of the defense in depth, which is the basic principle in safety, and referring to information related to domestic/foreign serious accidents, JNES conducted the fundamental study and made the following suggestions: 1) Definition of the serious accidents of the facilities. The definition is expected to contribute the discussion on new Rules by Nuclear Regulation Authority. 2) Methodology to examine the

  8. 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)

    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)

  9. Turning Atucha-1 PHWR from natural to slightly enriched uranium fuel operation: reassessing power limits for the first phase of transition

    International Nuclear Information System (INIS)

    Main features of Atucha-1 PHWR design and actual operation, using natural Uranium fuel, are outlined. The low exit burnup of this fuel, with its impact on the electricity generating cost, led to the implementation of a major change in the operation of this rather aged NPP, namely, the use of slightly enriched Uranium fuel. The original safety limits established for natural Uranium operation were focused on limits to power values for the central core region, with less well defined limits for the rest of the core, where the new fuel assemblies will soon be loaded. Recently, a number of assumptions were adopted and thermal-hydraulic calculations with Cobra-4 code were carried out in order to established consistent power limits for all of the core's radial hydraulic zones under different operating conditions. Finally, other margins were included in order to link the thermal-hydraulic limits to the final power limits to be applied in neutronic calculations, either for operation follow-up or in-core fuel management. (author)

  10. 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.

  11. Quaternion Averaging

    Science.gov (United States)

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

    2007-01-01

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

  12. Promotion of uranium enrichment business

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

  14. 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

    2015-01-01

    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.

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

    International Nuclear Information System (INIS)

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

  16. Study on possible usage of low enriched UO2 TVR-S type fuel elements for design of an experimental ADS research reactor

    International Nuclear Information System (INIS)

    Conceptual design of an accelerator driven sub-critical research reactor (ADSRR) was initiated at the Vinca Institute in 1999. Neutronic studies of the ADSRR-H were carried out for available high-enriched uranium (HEU) dioxide dispersed type TVR-S fuel elements (FEs) in lead matrix. Proton or deuteron beam would be extracted from the TESLA accelerator. In compliance with the Reduced Enrichment for Research and Test Reactors Program (RERTR), usage of HEU FEs in research reactors is not further recommended. Vinca Institute has returned fresh HEU FEs back to the Russia in 2002. New conceptual design of an ADSRR-L is based on assumed availability of low-enriched uranium dispersed type TVR-S FEs. Numerical simulations, carried out by Monte Carlo codes, show that neutron spectrum characteristics of ADSRR-L are compared to ones of the ADSRR-H with the same mass of 235U nuclide per FE and similar to well-known lead moderated and cooled power ADS with intermediate neutron spectrum. (author)

  17. Average Interest

    OpenAIRE

    George Chacko; Sanjiv Ranjan Das

    1997-01-01

    We develop analytic pricing models for options on averages by means of a state-space expansion method. These models augment the class of Asian options to markets where the underlying traded variable follows a mean-reverting process. The approach builds from the digital Asian option on the average and enables pricing of standard Asian calls and puts, caps and floors, as well as other exotica. The models may be used (i) to hedge long period interest rate risk cheaply, (ii) to hedge event risk (...

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

    OpenAIRE

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    牛淼淼; 黄亚继; 金保昇; 王妍艳; 董新新

    2014-01-01

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

  9. 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)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-16

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

  11. Fuel assembly

    International Nuclear Information System (INIS)

    The cross section of a fuel assembly is divided to a first region containing corner portions at which channel fasteners are situated and a second region not containing corner portions. The average enrichment degree of plutonium in the first region is decreased than that of the second region, and the number of fuel rods containing burnable poisons is increased at the first region than that of the second region. In the first region of the fuel assembly, the effect of moderating neutrons is enhanced since the cross section of a moderator flow channel at the outer side of the channel box is large. Therefore, local power peaking is increased in the first region while it is decreased in the second region that opposes to a narrow gap. The average enrichment degree of plutonium in the first region is decreased and that in the second region is increased by so much, to flatten the power distribution. Then, the reduction of the reactivity worth of gadolinia, as burnable poisons, can be suppressed. (N.H.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-15

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

  13. Enriched Gadolinium as burnable absorber for PWR

    International Nuclear Information System (INIS)

    This paper is a summary of a master of thesis work in reactor physics made by Ola Seveborn. The work was done at Vattenfall Braensle AB and Ola was guided through the work by the corresponding author of this paper. The results presented are calculations for Ringhals 3, which is a Westinghouse 3-loop PWR within the Vattenfall Group. The fuel is characterized by 17x17 assemblies of AFA type containing 3.80-3.95 w/o 235U and 8 rods containing 2 w/o Gadolinium with an enrichment of 70 w/o 157Gd. The calculations were performed with the Studsvik-Scandpower code package based on the CASMO-4 lattice code and the SIMULATE-3 nodal code. The results are compared to the corresponding calculations for fuel with 5 w/o gadolinium with natural isotopic constitution. The depletion of the cores was done separately for the reference and enriched case. The results show that the gains in average for the five cycles studied are about 70 EFPH per cycle. This is an effect of the lower gadolinium content needed. Also less parasitic absorption of enriched gadolinium in the end of the fuel life contributes to the increased cycle lengths. The abruptly increased reactivity and internal power peaking factor around 10 MWd/kgU do not affect the core design negatively. (authors)

  14. Uranium enrichment (a strategy analysis overview)

    International Nuclear Information System (INIS)

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

  15. Using the second law of thermodynamics for enrichment and isolation of microorganisms to produce fuel alcohols or hydrocarbons.

    Science.gov (United States)

    Kohn, Richard A; Kim, Seon-Woo

    2015-10-01

    Fermentation of crops, waste biomass, or gases has been proposed as a means to produce desired chemicals and renewable fuels. The second law of thermodynamics has been shown to determine the net direction of metabolite flow in fermentation processes. In this article, we describe a process to isolate and direct the evolution of microorganisms that convert cellulosic biomass or gaseous CO2 and H2 to biofuels such as ethanol, 1-butanol, butane, or hexane (among others). Mathematical models of fermentation elucidated sets of conditions that thermodynamically favor synthesis of desired products. When these conditions were applied to mixed cultures from the rumen of a cow, bacteria that produced alcohols or alkanes were isolated. The examples demonstrate the first use of thermodynamic analysis to isolate bacteria and control fermentation processes for biofuel production among other uses. PMID:26231417

  16. Blueprint for domestic uranium enrichment

    International Nuclear Information System (INIS)

    The AEC advisory committee on domestic production of uranium enrichment has studied for more than a year how to achieve the domestic enrichment of uranium by the construction and operation of a commercial enriching plant using centrifugal separation method, and the report was submitted to the Atomic Energy Commission on August 18, 1980. Japan has depended wholly on overseas services for her uranium enrichment needs, but the development of domestic enrichment has been carried on in parallel. The AEC decided to construct a uranium enrichment pilot plant using centrifuges, and it has been forwarded as a national project. The plant is operated by the Power Reactor and Nuclear Fuel Development Corp. since 1979. The capacity of the plant will be raised to approximately 75 ton SWU a year. The centrifuges already operated have provided the first delivery of fuel of about 1 ton for the ATR ''Fugen''. The demand-supply balance of uranium enrichment service, the significance of the domestic enrichment of uranium, the evaluation of uranium enrichment technology, the target for domestic enrichment plan, the measures to promote domestic uranium enrichment, and the promotion of the construction of a demonstration plant are reported. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-27

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

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

    International Nuclear Information System (INIS)

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

  1. Is Job Enrichment Really Enriching?

    OpenAIRE

    Robert D. Mohr; Cindy Zoghi

    2006-01-01

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

  2. Uranium enrichment

    International Nuclear Information System (INIS)

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

  3. Comparison of DUPIC fuel composition heterogeneity control methods

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ko, Won Il [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

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

  4. Fuel Cycle of VVER-1000: technical and economic aspects

    International Nuclear Information System (INIS)

    The paper contains estimations of dependences of technical and economic characteristics of VVER-1000 fuel cycle on number of charged FAs and their enrichment. In the study following restrictions were used: minimum quantity of loaded fresh FAs is equal 36 FAs, a maximum one - 78 (79) FAs and fuel enrichment is limited by value 4,95 %. The following technical and economic characteristics are discussed: cycle length, average burnup of spent fuel, specific consumption of natural uranium, specific quantity of separative work, annual production of thermal energy, fuel component of electrical energy cost, electricity generation cost. Results of estimations are presented as dependences of researched characteristics on cycle length, quantity of loaded FAs and their enrichments. The presented information allows to show tendencies and ranges of technical and economic characteristics at change of fuel cycle parameters. This information can be useful for definition of the fuel cycle parameters which satisfy the requirements of power system and exploiting organizations. (authors)

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

    International Nuclear Information System (INIS)

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

  6. Oxygen enrichment incineration

    International Nuclear Information System (INIS)

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

  7. 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.

  8. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. VVER-440 fuel cycles possibilities using modified FA design

    International Nuclear Information System (INIS)

    A nearly equilibrium five-year cycle has been achieved at Dukovany NPP in the last years. This means that working fuel assemblies (WFA) with an average enrichment of 4.25 w% of 235U (control assemblies with an average enrichment of 3.82 w% of 235U) are normally loaded and reloaded for five years. Operation at up rated thermal power (105% of the original one, increase from 1375 MWth to 1444 MWth) started by use of WFA with an average enrichment of 4.38 w% of 235U (control assemblies with an average enrichment of 4.25 w% of 235U) in 2009. With the aim of fuel cycle economy improvement, the fuel residence time in the core has to be prolonged up to six years with one cycle duration time and preserving loadings with very low leakage. In order to achieve this goal, at least neutron-physical characteristics of fuel assemblies must be improved and such changes should be evaluated from other viewpoints. Some particular changes have already been analyzed earlier. Designs of new fuel assemblies with higher (and in the central part of a fuel assemblies the highest possible, i.e. 4.95 w% of 235U) enrichment with preserving low pin power non-uniformity are described in the presented paper. An FA with an average enrichment of 4.76 w% of 235U (lower than originally evaluated) containing six fuel pins with 3.35 w% of Gd2O3 content was selected in the end. Fuel pins have bigger pellet diameter, but preserved central hole. A newly designed fuel assemblies were evaluated at first from the viewpoint of physics (pin power nonuniformity, cycle length etc.). Possibilities of fuel cycles are evaluated on model loadings with the newly designed fuel assemblies, where the base are loadings for twenty seventh-thirty fourth cycles of the third unit of Dukovany NPP for up rated power. These cycles were prolonged (from approx 330 FPD to 370 FPD) using fuel assemblies with higher enrichment. Also, a preliminary evaluation of fuel assemblies with a quite new design is presented. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Wenzel, Thomas P

    2009-10-27

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

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

    International Nuclear Information System (INIS)

    CERCA (Compagnie pour l'Etude et la Realisation de Combustibles Atomiques) is a private French Company that was set up more than 20 years ago, in 1957. The head office of our Company is located in Paris. We have an industrial center at Bonneuil-sur-Marne near Paris, and a research and production center at Romans, between the cities of Lyon and Marseille in southern France. Throughout its existence, CERCA has many times brought a significant contribution to the design and manufacturing procedures of fuels of all kinds: - graphite-gas - heavy water, gas cooled (El-4) - heavy water, water cooled (El-3 previous tube and snow crystal designs) - light water - high temperatures and also sophisticated control rods, fuel followers, etc. Among a wide variety of other type of fuel elements CERCA has been involved since 1960 in MTR fuel element production and operates in Romans a production unit capable of 15 to 20,000 fuel plates per year. The backing of our strong R and D department enabled not only to follow the regularly increasing demand for improved characteristics by the reactor designers and users, but also to remain always a little bit ahead of these requirements in order to be able at all times to face new demands. The present international concern about non-proliferation of weapons-grade enriched uranium is now focusing interest on fuel elements with high total uranium contents. The current performance level of our products, that was made possible by our previous R and D programs, is such that we will experience no problems in keeping many low or medium power research reactors in operation. About one year ago, CERCA decided to initiate a program of technological development, to face the problem of the reduction of available enrichment

  14. Job Enrichment

    Science.gov (United States)

    Sanders, Rick

    1970-01-01

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

  15. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

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

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To improve the thermal and mechanical safety of fuel rods and structural components by making the local power coefficient of jointed fuel rods greater than that of other fuel rods in a fuel assembly. Constitution: In a fuel assembly comprising a plurality of fuel rods bundled by a spacer and held at the upper and the lower positions with tie plates for insertion into a channel, the degree of enrichment of uranium 235 for uranium dioxide fuel pellets charged in jointed fuel rods is adjusted such that the local power coefficient of the jointed fuel rods is made greater than that of the other fuel rods. In the case if the upper tie plate is moved upwardly by the extension of the jointed fuel rods, other fuel rods axially free from the upper tie plate receives no tension, whereby the safety of the fuel assembly can be improved. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

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

  18. Advances in uranium enrichment processes

    International Nuclear Information System (INIS)

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

  19. Fuel burnup characteristics for the NRU research reactor

    International Nuclear Information System (INIS)

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U3Si, consisting of particles of U3Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  20. Fuel burnup characteristics for the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U{sub 3}Si, consisting of particles of U{sub 3}Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

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

    International Nuclear Information System (INIS)

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

  2. Uranium enrichment

    International Nuclear Information System (INIS)

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

  3. Uranium enrichment

    International Nuclear Information System (INIS)

    This chapter discusses the development of uranium enrichment processes. In the introduction there is a brief history of uranium enrichment, followed by a summary of the criteria used for the assessment of an isotope separation process, e.g. the separation factor, separative power, and the power consumption of a separating element. This is followed by a discussion of the two main processes used, i.e. gaseous diffusion and centrifugation. The reason for the change from diffusion to centrifugation in the UK, mainly on power costs, is discussed. The development potential of centrifuges is also assessed. Other processes which have been developed up to pilot stage are described, e.g. the Becker jet nozzle and the South African process. This is followed by a description of some plasma-based methods. The next topic is concerned with chemical exchange methods and an attempt is made to assess their potential in the enrichment scene from published information. This chapter concludes with a discussion of the advanced laser isotope-separation methods. The two approaches, i.e. the atomic and the molecular routes are discussed again using published information. This information is insufficient to give a complete assessment of the methods, especially the molecular route, but is enough to give indications of their potential

  4. Enriching recycled uranium

    International Nuclear Information System (INIS)

    The paper reviews the progress of the use of recycled uranium during the period 1985-8. This article was originally presented as a paper at the 1988 Uranium Institute symposium (which was held in London). A description is given of the differences between natural and recycled uranium, and the presence of U236 in recycled uranium. The concept of equivalent reactivity is described, as well as the cost benefit of using recycled uranium. A summary of Urenco tests and trials with reprocessed uranium is given. Enrichment, UF6 conversion and fuel fabrication are also discussed. (U.K.)

  5. WWER-440 fuel cycles possibilities using modified fuel assemblies design

    International Nuclear Information System (INIS)

    A nearly equilibrium five-year cycle has been achieved at Dukovany NPP over the last years. This means that working fuel assemblies with an average enrichment of 4.25 w % (control assemblies) with an average enrichment of 3.82 w %) are normally loaded and reloaded for five years. Operation at uprated thermal power (105% of the original one, increase from 1375 MWt to 1444 MWt) is being prepared by use of working fuel assemblies with an average enrichment of 4.38 w % (control assemblies with an average enrichment of 4.25 w %). With the aim of fuel cycle economy improvement, the fuel residence time in the core has to be prolonged up to six years with one cycle duration time up to 18 months and preserving loadings with very low leakage. In order to achieve this goal, at least neutron-physical characteristics of fuel assemblies must be improved and such changes should be evaluated from other viewpoints. Some particular changes have already been analyzed earlier. Designs of new fuel assemblies with higher (and in the central part of a fuel assemblies the highest possible, i.e. 4.95 w %) enrichment with preserving low pin power non-uniformity are described in the presented paper. An fuel assemblies with an average enrichment of 4.66 w % (lower than originally evaluated) containing six fuel pins with 3.35 w % Gd2O3 content was selected in the end. Fuel pins have bigger pellet diameter, bigger pin pitch and thinner fuel assemblies shroud. A newly designed fuel assemblies was evaluated from the viewpoint of physics (pin power non-uniformity, criticality of fuel at transport and storage and determination of basic quantities for spent fuel storage purposes by ORIGEN code), thermo-hydraulics (comparison of subchannel output temperatures and the departure from nucleate boiling ratio - DNBR) and mechanical properties. The purpose of this study was to simulate an fuel assemblies subject to the loads during its six- year lifetime whereas normal working conditions were taken into

  6. Uranium enrichment: technology, economics, capacity

    International Nuclear Information System (INIS)

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

  7. An evaluation of the VM/VF ratio to standard UO2 and MOX fuel with 4,5% enrichment and 1% of neptunium insertion

    International Nuclear Information System (INIS)

    Many ideas are being studied in order to burn up the high radioactivity wastes, including the insertion of minor actinides in reactors. Some indicate the possibility of differentiated burnup when studying different VM/VF. The VM/VF ratio, moderator volume/fuel volume, is directly related with the value obtained for the multiplication factor k. There is a VM/VF which k is maximum, and this is directly related with the fuel composition. This work is a study, using WIMS-D5 code, to find the better value of VM/VF, considering MOX and UO2 fuels with 1% Neptunium insertion. The MOX is a fuel reprocessed by PUREX technical. The following parameters are evaluated: spectrum hardening, boron effect, and the fuel and moderator temperature coefficients. (author)

  8. Heterogeneous reactivity effects in medium- and high-enriched uranium metal-water systems

    International Nuclear Information System (INIS)

    The effect of heterogeneity on reactivity of low-, medium-, and high-enriched, water-moderated uranium metal systems has been examined for various hydrogen-to-fissile (H/X) ratios using the CSAS1X sequence in SCALE and MCNP. For the calculations, an infinite array of close-packed unit cells was modeled which consisted of centered uranium metal spheres surrounded by water. The enrichments used correspond to the average enrichments of fragmented fuel plates in three proposed waste shipments from Oak Ridge National Laboratory. The analysis performed to obtain peak reactivity for each enrichment as a function of particle size and H/X ratio led to the development of the topic discussed in this paper

  9. Enrichment reduction for research reactors

    International Nuclear Information System (INIS)

    The worldwide activities on enrichment reduction for research reactors are reviewed and the national and international programs are described. Especially the following points are discussed: Benchmark calculations, reactor safety, fuel element development, irradiation tests, post irradiation examinations, full core demonstrations, activities of the GKSS and economical questions. (orig.)

  10. Fuel burnup extension effect on the fuel utilization and economical impact for a typical PWR plant

    International Nuclear Information System (INIS)

    Currently in Japan, fuel assembly average burn-up is limited to 48GWd/t and is going to be extended to 55GWd/t in these years. Moreover, R and D programs for further extension are under operation. Simultaneous extension of fuel burn-up limitation and cycle length reduces the number of fuel required to produce a given amount of energy reducing the radioactive waste generation, the occupational radiation exposure and the electricity generation cost. In this paper, the effect of fuel burn-up and operation cycle length extension is estimated from the view point of electricity generation cost and amount of discharged fuel assemblies, and the desirable burn-up extension in the future is studied. The present 5wt% uranium-235 enrichment restriction for commercial reactors divides the burn-up extension implementation in two steps. The fuel burn-up achievable with the present 5wt% enrichment limitation and without it is analyzed. A standard 3 loop PWR plant loading 17x17 fuel assemblies has been chosen for the feasibility study of operation cycle longer than 15 months and up to 24 months under extended fuel burn-up limitation. With the 5wt% enrichment limitation, the maximum assembly average burn-up is between 60GWd/t and 70GWd/t. Three batches reload fuel strategy and 18 months operation cycle allow the electricity generation cost reduction in about 4% and the number of fuel assemblies discharged per year is reduced in approximately 15% compared with the current 48GWd/t fuel. Relaxing the enrichment limitation, for the 24 months operation cycle with 3 batches reload fuel strategy, the maximum assembly average burn-up become 80GWd/t. The electricity generation cost reduction is about 8% and the number of fuel assemblies discharged per year is reduced in approximately 35% compared with the current condition. This study shows the contribution of simultaneous extension of fuel burn-up limitation and operation cycle length to reduce the electricity generation cost and the number

  11. Synergistic CANDU-LWR fuel cycles

    International Nuclear Information System (INIS)

    CANDU is the most neutron-efficient reactor available commercially, allowing utilization of a range of fuel cycles. The flexibility of on-line refuelling allows fuel management to accommodate these different fuels. A synergism with light-water reactors (LWR) is possible through the use in CANDU of uranium and/or plutonium recovered from spent LWR fuel. In the TANDEM fuel cycle, the unseparated uranium and plutonium (1.5% fissile) would give a burnup in CANDU of about 25 MW.d/kg HE, producing four times more energy than that available from simply recycling the plutonium in an LWR. In another potential fuel cycle, uranium recovered from spent LWR fuel during conventional reprocessing is also recycled in CANDU, without re-enrichment. An average recovered uranium (RU) enrichment of 0.9% in U-235 results in a CANDU burnup of at least 13 MW.d/kg U, allowing twice as much energy to be extracted, compared with that from an LWR. The fuelling cost for RU in CANDU are about 35% lower than for natural uranium. Additionally, direct use of spent LWR fuel in CANDU is theoretically possible, but requires practical demonstration. AECL and KAERI are developing the CANFLEX (CANDU Flexible Fuelling) advanced fuel bundle as the optimal carrier for all extended burnup fuel cycles envisaged for CANDU

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

    OpenAIRE

    Olalekan Wasiu Saheed; Rashid A.A.; Baharom Masri

    2014-01-01

    Since the pressure development in a combustion chamber is uniquely related to the combustion process, substantial variations in the combustion process on a cycle-by-cycle basis are occurring. To this end, an experimental study of cycle-by-cycle variation in a direct injection spark ignition engine fueled with natural gas-hydrogen blends combined with exhaust gas recirculation at relative air-fuel ratios was conducted. The impacts of relative air-fuel ratios (i.e. λ = 1.0, 1.2, 1.3 and 1.4 whi...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  14. The enrichment industry reaches maturity

    International Nuclear Information System (INIS)

    As the nuclear power industry enters the 1980s, uranium enrichment supply can no longer be considered one of the critical problem areas of the nuclear fuel cycle. It has become an industrial and commercial activity which has reached a high degree of maturity. Three main aspects of this maturity are discussed: 1. the availability of enrichment services from several facilities with very diverse ownership; 2. the involvement of private industry, especially in Europe, and the application of normal commercial rules to enrichment contracts; 3. the ability of the enrichment industry to cope with recent setbacks in the advancement of nuclear power programmes whilst carrying out an active research and development programme that will help to ensure its future technical and economic viability. (U.K.)

  15. An evaluation of the VM/VF ratio to standard UO2 and MOX fuel with 4,5% enrichment and 1% of americium insertion

    International Nuclear Information System (INIS)

    A growing interest exists in the development of techniques for burning and transmuting minor actinides. Some indicate the possibility of differentiated burnup when studying different VM/VF. The VM/VF ratio, moderator volume/fuel volume, is directly related with the value obtained for the multiplication factor k. There is a VM/VF for which k is maximum, and this is directly related with the fuel composition. This work is a study to find a better value of VM/VF, using the WIMS-D5 code, considering a UO2 fuel and a MOX fuel, with 1% Americium insertion. The following parameters were appraised: spectrum hardening, boron worth, and reactivity temperature coefficients. (author)

  16. 75 FR 62895 - Notice of Availability of Safety Evaluation Report; AREVA Enrichment Services LLC, Eagle Rock...

    Science.gov (United States)

    2010-10-13

    ... Project Manager, Advanced Fuel Cycle, Enrichment, and Uranium Conversion, Division of Fuel Cycle Safety... material in a gas centrifuge uranium enrichment facility. The applicant proposes that the facility, known..., decommissioning, management measures, physical protection, and materials control and accountability. III....

  17. Nuclear fuel storage

    International Nuclear Information System (INIS)

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

  18. Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U3O8, UAl2 and U3Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the REM Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U3Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup. (author)

  19. Bio-CCS: co-firing of established greenfield and novel, brownfield biomass resources under air, oxygen-enriched air and oxy-fuel conditions

    OpenAIRE

    Pickard, S; Daood, SS; Nimmo, W.; Lord, R; Pourkashanian, M.

    2013-01-01

    As demand for electricity and atmospheric CO concentrations rise technologies that reduce the environmental impact of generating electricity are sought. Within the many options a combination of co-firing of biomass and carbon capture and storage (Bio-CCS) could present a negative-emission process. This work investigates co-firing of a novel brownfield and two conventional greenfield biomass reserves with coal in oxygen-enriched conditions which may enhance the efficiency of post-combustion ca...

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

    Directory of Open Access Journals (Sweden)

    Olalekan Wasiu Saheed

    2014-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Graven, H. D.; Gruber, N.

    2011-01-01

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

  5. Contribution of CERCA to the US DOE conference on the use of 20% and 45% enriched uranium as fuel for research reactions. Part 1

    International Nuclear Information System (INIS)

    Among a wide variety of other type of fuel elements CERCA has been involved since 1960 in MTR fuel element production and operates in Romans a production unit capable of 15 to 20,000 fuel plates per year. The backing of our strong R and D department has made it possible, not only to follow the regularly increasing demand for improved characteristics by the reactor designers and users, but also to remain always a little bit ahead of these requirements in order to be able at all times to face new demands. Starting with the now nearly obsolete uranium-aluminum alloy picture-frame technique, we soon had to develop powder metallurgy techniques, in parallel with improved bonding techniques adapted to the harder aluminum alloy cladding that becomes necessary in order to keep up with the increasing uranium content of the fuel plates. Of course the main concern at each new step of technical development is the dogbone and the homogeneity specifications along with the necessity of maintaining high fabrication yields in order to keep the costs as low as possible. At each step, irradiation experience was obtained in the French CEA reactors

  6. Status of the development of RU-43 fuel at INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G. [Institute of Nuclear Research, Pitesti (Romania)

    2008-07-01

    More than 50000 fuel bundles containing natural uranium fuel have been irradiated in the CANDU-6 reactors of Cernavoda-Romania NPP, with a very low defect rate, to a core-average discharge burnup of 170-190 Mwh/kgU. Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. After fission products and plutonium (Pu) have been removed from spent LWR fuel, RU is left. A fissile content in the RU of 0.9 to 1.1% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. RU from spent LWR fuel can be considered as a lower cost source of enrichment at the optimal enrichment level for CANDU fuel pellets. In Europe the feedstock of RU is approaching thousands tones and would provide sufficient fuel for hundreds CANDU-6 reactors years of operation. The use of RU fuel offers significant benefits to CANDU reactor operators. RU fuels improves fuel cycle economics by increasing the fuel burnup, which enables large cost reductions in fuel consumption and in spent fuel disposal. RU fuel offers enhanced operating margins that can be applied to increase reactor power. These benefits can be realized using existing fuel production technologies and practices, and with almost negligible changes to fuel receipt and handling procedures at the reactor. The application of RU fuel could be an important element in Cernavoda NPP. For this reason the Institute for Nuclear Research (INR), Pitesti has started a research programme aiming to develop a new fuel bundle RU-43 for extended burnup operation. The most relevant calculations performed on this fuel bundle design version are presented. Also, the stages of an experimental program aiming to verify the operating performance are briefly described in this paper. (orig.)

  7. Fuel cycle data survey

    International Nuclear Information System (INIS)

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

  8. Neutron resonance averaging

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. 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)

    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

  11. Averaging anisotropic cosmologies

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Contribution of CERCA to the US DOE conference on the use of 20% and 45% enriched uranium as fuel for research reactions. Part 2

    International Nuclear Information System (INIS)

    As far as prices are concerned, the author would like to state first that for the time being we have to compare the prices of a steady state situation, which is represented by a well known process that has been in use for many years, with a transitory situation which is the achievement of the same expertise in new, extrapolated fuels. That is why the comparison has to be corrected for the results within the next 2 to 3 years. Obviously, it seems that there are other factors which contribute to the price increases. I think there is a very significant example for this

  14. Update on international uranium and enrichment supply

    International Nuclear Information System (INIS)

    Commercial nuclear power generation came upon us in the late 1950s and should have been relatively uneventful due to its similarities to fossil-powered electrical generation. Procurement of nuclear fuel appears to have been treated totally different from the procurement of fossil fuel, however, and only recently have these practices started to change. The degree of utility reliance on US-mined uranium and US Dept. of Energy (DOE)-produced enrichment services has changed since the 1970s as federal government uncertainty, international fuel market opportunity, and public service commission scrutiny has increased. Accordingly, the uranium and enrichment market has recognized that it is international just like the fossil fuel market. There is now oversupply-driven competition in the international nuclear fuel market. Competition is increasing daily, as third-world countries develop their own nuclear resources. American utilities are now diversifying their fuel supply arrangements, as they do with their oil, coal, and gas supply. The degree of foreign fuel arrangements depends on each utility's risk posture and commitment to long-term contracts. In an era of rising capital, retrofit, operating, and maintenance costs, economical nuclear fuel supply is even more important. This economic advantage, however, may be nullified by congressional and judicial actions limiting uranium importation and access to foreign enrichment. Such artificial trade barriers will only defeat US nuclear generation and the US nuclear fuel industry in the long term

  15. Nuclear fuel

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    张朝山; 熊树生; 任晓帅; 姚红; 徐进; 谢莲; 刘震涛

    2012-01-01

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

  18. The ORR Whole-Core LEU Fuel Demonstration

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Average-energy games

    OpenAIRE

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

    2015-01-01

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

  2. Reprocessing RERTR silicide fuels

    International Nuclear Information System (INIS)

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High-density, low-enrichment, aluminum-clad uranium silicide fuels may be substituted for the highly enriched aluminum-clad alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR silicide fuels at Savannah River Plant. Results of dissolution and feed preparation tests and solvent extraction processing demonstrations with both unirradiated and irradiated uranium silicide fuels are presented

  3. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

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

  4. Average Angular Velocity

    OpenAIRE

    Van Essen, H.

    2004-01-01

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

  5. On the Averaging Principle

    OpenAIRE

    Fibich, Gadi; Gavious, Arieh; Solan, Eilon

    2012-01-01

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

  6. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

    1999-01-01

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

  7. Averaged extreme regression quantile

    OpenAIRE

    Jureckova, Jana

    2015-01-01

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

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Fuel rods enriched with plutonium and fuel rods formed by incorporating combustible poisons in enriched uranium are arranged in square lattice like structure. MOX fuel pellets comprise PuO2 as a fuel material and contain 239Pu, 241Pu as fission products. The gadolinia-incorporated uranium fuel pellets comprise UO2 as a fuel material and gadolinia as a burnable poison incorporated therein and contains 235U as a fuel material. The axial distribution of the concentration of gadolinia contained in the uranium fuel rods is axially divided into three regions in a region less than 1/2 of a fuel effective length, and the concentration of gadolinia is highest at the lowest region, and the concentration of gadolinia is made lower toward the upper regions. With such a constitution, the degree of downward distortion of the axial power distribution is suppressed in a reactor core of a BWR type reactor having a large MOX loading rate. (I.N.)

  9. Fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Izutsu, Sadayuki; Fujita, Satoshi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Fujimaki, Shingo; Sasagawa, Masaru; Kaneto, Kunikazu; Mochida, Takaaki; Aoyama, Motoo; Shimada, Hidemitsu

    1997-09-09

    Fuel rods enriched with plutonium and fuel rods formed by incorporating combustible poisons in enriched uranium are arranged in square lattice like structure. MOX fuel pellets comprise PuO{sub 2} as a fuel material and contain {sup 239}Pu, {sup 241}Pu as fission products. The gadolinia-incorporated uranium fuel pellets comprise UO{sub 2} as a fuel material and gadolinia as a burnable poison incorporated therein and contains {sup 235}U as a fuel material. The axial distribution of the concentration of gadolinia contained in the uranium fuel rods is axially divided into three regions in a region less than 1/2 of a fuel effective length, and the concentration of gadolinia is highest at the lowest region, and the concentration of gadolinia is made lower toward the upper regions. With such a constitution, the degree of downward distortion of the axial power distribution is suppressed in a reactor core of a BWR type reactor having a large MOX loading rate. (I.N.)

  10. Conceptual design of KALIMER uranium metallic fueled core

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time

  13. Uranium enrichment capacity: public versus private ownership

    International Nuclear Information System (INIS)

    Continual growth of conventional nuclear capacity requires an assured supply of enriched uranium and, hence, potential expansion of domestic uranium enrichment capacity. The question of ownership of new enrichment capacity, i.e., public or private, entails not only the social-opportunity costs of alternative investments but also technical parameters of uranium utilization and advanced reactor development. Inclusion of risk preferences in both the public and private sectors produces interesting results in terms of optimal investment strategies with respect to choice of technology and scale of investment. Utilization of a nuclear fuel cycle requirements process model allows explicit specification of production technology. Integration of process model output with a least-cost investment model permits flexibility in parametric analysis. Results indicate minimum incentive for Government subsidy of a private enrichment sector through 2000 given moderate to low nuclear growth assumptions. The long-run scenario, to 2020, exhibits potentially greater incentives for private enrichment investment

  14. Beyond Job Enrichment to Employment Enrichment

    Science.gov (United States)

    Werther, William B., Jr.

    1975-01-01

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

  15. Averaging anisotropic cosmologies

    CERN Document Server

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

    2006-01-01

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

  16. Average Angular Velocity

    CERN Document Server

    Essén, H

    2003-01-01

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

  17. On sparsity averaging

    CERN Document Server

    Carrillo, Rafael E; Wiaux, Yves

    2013-01-01

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

  18. 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.

  19. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

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

  20. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

    2001-01-01

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

  1. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  2. Covariant approximation averaging

    CERN Document Server

    Shintani, Eigo; Blum, Thomas; Izubuchi, Taku; Jung, Chulwoo; Lehner, Christoph

    2014-01-01

    We present a new class of statistical error reduction techniques for Monte-Carlo simulations. Using covariant symmetries, we show that correlation functions can be constructed from inexpensive approximations without introducing any systematic bias in the final result. We introduce a new class of covariant approximation averaging techniques, known as all-mode averaging (AMA), in which the approximation takes account of contributions of all eigenmodes through the inverse of the Dirac operator computed from the conjugate gradient method with a relaxed stopping condition. In this paper we compare the performance and computational cost of our new method with traditional methods using correlation functions and masses of the pion, nucleon, and vector meson in $N_f=2+1$ lattice QCD using domain-wall fermions. This comparison indicates that AMA significantly reduces statistical errors in Monte-Carlo calculations over conventional methods for the same cost.

  3. The averaging principle

    OpenAIRE

    Fibich, Gadi; Gavious, Arieh; Solan, Eilon

    2012-01-01

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

  4. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  5. DOE enrichment plant hums ahead

    International Nuclear Information System (INIS)

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

  6. Use of slightly enriched uranium (SEU) in PHWR

    International Nuclear Information System (INIS)

    The flexibility of CANDU reactor allows the use of different fissile materials: Natural Uranium (NU), Slightly Enriched Uranium (SEU), Recovered Uranium (RU) from LWR spent fuel reprocessing, thorium or plutonium (Mixed Oxide - MOX) and even, PWR spent fuel (direct use of spent fuel in CANDU - DUPIC). This is possible using as carrier in the pressure tube an advanced fuel bundle. AECL Canada developed a new fuel bundle concept named CANFLEX (Canadian Flexible). In the irradiation demonstration at Point Lepreau Reactor, using CANFLEX fuel bundle with natural uranium, the burnup was about 10,000 MWd/tU comparatively with 7000 MWd/tU for standard fuel bundle with natural uranium. Slightly Enriched Uranium is uranium having the enrichment 0.9 - 1.2% 235U. There are estimations that the uranium enrichment increasing up to 1.2% 235U results in a burnup of about 21,000 MWd/tU. In Romania, also, Institute for Nuclear Research Pitesti develops a new fuel bundle concept (named SEU 43) compatible with CANDU 6 Reactor. The intention is to use slightly enriched Uranium for burnup increasing and to reduce spent fuel amounts. (author)

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

    Institute of Scientific and Technical Information of China (English)

    李延吉; 姜璐; 赵宁; 李玉龙; 李润东; 池涌

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

  10. Analysis of fuel options in TRIGA reactor

    International Nuclear Information System (INIS)

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

  11. Robust Averaging Level Control

    OpenAIRE

    Rosander, Peter; Isaksson, Alf; Löfberg, Johan; Forsman, Krister

    2011-01-01

    Frequent inlet flow changes typically cause problems for averaging level controllers. For a frequently changing inlet flow the upsets do not occur when the system is in steady state and the tank level at its set-point. For this reason the tuning of the level controller gets quite complicated, since not only the size of the upsets but also the time in between them relative to the hold up of the tank have to be considered. One way to obtain optimal flow filtering while directly accounting for futur...

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

    International Nuclear Information System (INIS)

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

  13. Negative Average Preference Utilitarianism

    Directory of Open Access Journals (Sweden)

    Roger Chao

    2012-03-01

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

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

    International Nuclear Information System (INIS)

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

  15. Average nuclear surface properties

    International Nuclear Information System (INIS)

    The definition of the nuclear surface energy is discussed for semi-infinite matter. This definition is extended also for the case that there is a neutron gas instead of vacuum on the one side of the plane surface. The calculations were performed with the Thomas-Fermi Model of Syler and Blanchard. The parameters of the interaction of this model were determined by a least squares fit to experimental masses. The quality of this fit is discussed with respect to nuclear masses and density distributions. The average surface properties were calculated for different particle asymmetry of the nucleon-matter ranging from symmetry beyond the neutron-drip line until the system no longer can maintain the surface boundary and becomes homogeneous. The results of the calculations are incorporated in the nuclear Droplet Model which then was fitted to experimental masses. (orig.)

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

    International Nuclear Information System (INIS)

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

  17. High Accuracy U-235 Enrichment Verification Station for Low Enriched Uranium Alloys

    International Nuclear Information System (INIS)

    The Y-12 National Security Complex is playing a role in the U.S. High Performance Research Reactor (USHPRR) Conversion program sponsored by the U.S. National Nuclear Security Administration's Office of Global Threat Reduction. The USHPRR program has a goal of converting remaining U.S. reactors that continue to use highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. The USHPRR program is currently developing a LEU Uranium-Molybdenum (U-Mo) monolithic fuel for use in the U.S. high performance research reactors.Y-12 is supporting both the fuel development and fuel fabrication efforts by fabricating low enriched U-Mo foils from its own source material for irradiation experiments and for optimizing the fabrication process in support of scaling up the process to a commercial production scale. Once the new fuel is qualified, Y-12 will produce and ship U-Mo coupons with verified 19.75% +0.2% - 0.3% U-235 enrichment to be fabricated into fuel elements for the USHPRRs. Considering this small enrichment tolerance and the transition into HEU being set strictly at 20% U-235, a characterization system with a measurement uncertainty of less than or equal to 0.1% in enrichment is desired to support customer requirements and minimize production costs. Typical uncertainty for most available characterization systems today is approximately 1-5%; therefore, a specialized system must be developed which results in a reduced measurement uncertainty. A potential system using a High-Purity Germanium (HPGe) detector has been procured, and tests have been conducted to verify its capabilities with regards to the requirements. Using four U-Mo enrichment standards fabricated with complete isotopic and chemical characterization, infinite thickness and peak-ratio enrichment measurement methods have been considered for use. As a result of inhomogeneity within the U-Mo samples, FRAM, an isotopic analysis software, has been selected for initial testing. A systematic approach

  18. Enrichment: Present and projected future supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Lenders, Maurice [URENCO, Buckinghamshire (United Kingdom)

    2009-04-15

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

  19. U.S. forms uranium enrichment corporation

    International Nuclear Information System (INIS)

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

  20. Advanced enrichment techniques

    International Nuclear Information System (INIS)

    BNFL is in a unique position in that it has commercial experience of diffusion enrichment, and of centrifuge enrichment through its associate company Urenco. In addition BNFL is developing laser enrichment techniques as part of a UK development programme in this area. The paper describes the development programme which led to the introduction of competitive centrifuge enrichment technology by Urenco and discusses the areas where improvements have and will continue to be made in the centrifuge process. It also describes the laser development programme currently being undertaken in the UK. The paper concludes by discussing the relative merits of the various methods of uranium enrichment, with particular reference to the enrichment market likely to obtain over the rest of the century. (author)

  1. Profile of World Uranium Enrichment Programs - 2007

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    2007-11-01

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

  2. Uranium enrichment: an overview

    International Nuclear Information System (INIS)

    This paper is a general presentation of uranium enrichment processes and assessments of the prevailing commercial and industrial situations. It gives first some theoretical aspects of enrichment in general and explains the differences between statistical and selective processes in particular. Then a review of the different processes is made with a comparison between them. Finally, some general remarks concerning applications are given and the risks of proliferation related to enrichment are mentioned. (J.S.). 4 refs., 5 figs., 8 tabs

  3. Enriched coalgebraic modal logic

    OpenAIRE

    Wilkinson, Toby

    2013-01-01

    We formalise the notion of enriched coalgebraic modal logic, and determine conditions on the category V (over which we enrich), that allow an enriched logical connection to be extended to a framework for enriched coalgebraic modal logic. Our framework uses V-functors L: A ? A and T: X ? X, where L determines the modalities of the resulting modal logics, and T determines the coalgebras that provide the semantics. We introduce the V-category Mod(A, ?) of models for an L-algebra (A, ?), and ...

  4. Fuel behavior comparison for a research reactor

    Science.gov (United States)

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

    2006-06-01

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

  5. Fuel behavior comparison for a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gh. [Institute for Nuclear Research (ICN), 1, Campului Street, P.O. Box 78, 0300 Mioveni, Pitesti (Romania)]. E-mail: joenegut@yahoo.com; Mladin, M. [Institute for Nuclear Research (ICN), 1, Campului Street, P.O. Box 78, 0300 Mioveni, Pitesti (Romania); Prisecaru, I. [University Politehnica Bucharest (Romania); Danila, N. [University Politehnica Bucharest (Romania)

    2006-06-30

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

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

    International Nuclear Information System (INIS)

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

  7. Improved type many kinds of enrichment initial charge core

    International Nuclear Information System (INIS)

    No. 4 plant in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc. started the commercial operation in September, 1993 after 48 months of construction period. It is the largest BWR plant with 1137 MWe and 3293 MWt output. The basic specifications of the plant and the core fuel are shown. In the design of the core fuel, the heightening of safety and reliability has priority, and the improved technology for bettering the core performance and operability was adopted positively. As the core fuel technology, new 8 x 8 zirconium-lined fuel, improved type many kinds of enrichment initial charge core, long life hafnium type control rods and so on were adopted. The zirconium-lined fuel is to reduce pellet-clad interaction, and increase the reliability. Moreover, the nuclear design of high reactivity fuel assemblies using uranium-saving technology was adopted. The core is composed of the fuels of three kinds of enrichment, and high enrichment fuel is arranged outside, thus the taking-out burnup of initial charge fuel increases. The improvement of the design of initial charge core, the improved type many kinds of enrichment initial charge core and the core characteristics in the start-up operation are reported. It was confirmed that the good core characteristics as expected can be obtained. (K.I.)

  8. Fast calculation program for nuclear fuel lattice design of boiling water reactors

    International Nuclear Information System (INIS)

    A methodology for 10 x 10 fuel lattice design for BWR's was developed. A linear equation based on a matrix of coefficients (change ratio of relative power due to U-235 enrichment changes) was proposed for estimating pin by pin relative powers of a fuel lattice. Based on above description the fast calculation program PreDiCeldas was developed which allows diminishing the maximum Local Power peaking Factor (LPPF). This program uses a simple search algorithm varying the distribution of U-235 within the fuel lattice, and also maintaining its average enrichment as a constant. The pin relative power estimation accuracy is of the order of 0.04% with respect to HELIOS transport code calculations. With PreDiCeldas program it was possible, in a short calculation time, to redesign a reference fuel lattice diminishing the maximum LPPF, at the beginning of the life, from 1.405 to 1.225. (authors)

  9. Uranium enrichment. Principles

    International Nuclear Information System (INIS)

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

  10. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-30

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

  11. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    International Nuclear Information System (INIS)

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

  12. HTGR fuel and fuel cycle experience in the United States

    International Nuclear Information System (INIS)

    In the United States, fuel and fuel cycle development for High-Temperature Gas-Cooled Reactors (HTGR's) has been concentrated on variations of the uranium-thorium fuel cycle. The most efficient cycle utilizes highly enriched U-235 and bred U-233. A fuel cycle utilizing a lower enrichment of about 20% fissile in U-238 also performs well and offers a high degree of protection against proliferation of potential weapons materials. Operating experience in the Peach Bottom Unit 1 and Fort St. Vrain HTGR's has demonstrated very favorable retention of fission products and a high integrity of the fuel element assemblies. Capsule irradiation tests of 20%-enriched fuels for later reactor designs have shown equally good fuel performance. A comprehensive program for developing shipping, storage, and reprocessing technology for HTGR fuel cycles is being carried out cooperatively by the United States and the Federal Republic of Germany

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2016-01-01

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

  18. Candu fuel and fuel cycles

    International Nuclear Information System (INIS)

    reactor designs, allowing operation today on currently available fuels and switching to other fueling options as market conditions change. This establishes an important freedom from future resource constraints without depending on future commercialization of challenging and expensive technologies such as fast breeder reactors, yet, once these are commercially available, CANDU and fast breeder fuel cycles are complementary and can achieve a highly advantageous synergism. This paper examines the fuel cycle option which CANDU reactor technology can accommodate, including the use of slightly enriched uranium direct use of spent pressurized water reactor fuel in CANDU (dupic), burning recovered uranium, mixed plutonium and uranium oxides or actinides and the use of thorium based fuel cycles. These options provide CANDU reactors with the most flexible fuelling of any reactor type, which are readily adaptable to meeting future variations in energy markets, regardless of what these may be. (author)

  19. Start-up fuel and power flattening of sodium-cooled candle core

    International Nuclear Information System (INIS)

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

  20. Fabrication experience and integrity confirmation tests of the first-loading-fuel of the HTTR

    International Nuclear Information System (INIS)

    The first-loading core of the High Temperature Engineering Test Reactor (HTTR) consists of 150 fuel assemblies. An HTTR fuel assembly is so-called a pin-in-block type of hexagonal graphite block. A fuel rod consists of a graphite sleeve and of 14 fuel compacts. In a fuel compact, about 13,000 TRISO coated fuel panicles are dispersed densely. The coated fuel panicle is TRISO (Tri-isotropic) type with four coating layers. The fuel kernel is low-enriched (average 6wt%) UO2. The fabrication of the first-loading fuel started from June 1995. A total of 4,770 fuel rods were successfully produced and transferred to the reactor building of the HTTR. Finally, in the reactor building, the fuel rods were inserted to the graphite blocks to form fuel assemblies. On December 1997, 150 fuel assemblies were completely formed and were stored in new fuel storage cells. The cells were filled with helium gas to keep the fuel blocks in dry condition. Fabrication technology of the HTTR fuel was established through a lot of R and D activities and fabrication experiences of irradiation examination samples spread over about 30 years. High quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and optimization of the compaction conditions. In the safety design of HTGR fuel, it is important to retain fission products within the coated fuel panicles so that their release to the primary coolant may not exceed an acceptable level. From this point of view, as-fabricated failure fraction is important. In the specification, SiC-failure and exposed uranium fractions were determined to be less than 1.5x10-3 and l.5x10-4, respectively. The quality of the first loading fuel fully satisfied the design specifications for the fuel. The fuel compacts contained almost no through-coatings failed particles and few SiC-defective particles. Average through-coatings and SiC defective

  1. Uranium conversion and enrichment

    International Nuclear Information System (INIS)

    A description is given of the Atomic Energy Corporation's uranium conversion and enrichment plants at Valinda ba, including a brief discussion of problems encountered and plans for future developments. (author)

  2. Fuel Element Designs for Achieving High Burnups in 220 MW(e) Indian PHWRs

    International Nuclear Information System (INIS)

    Presently 19-element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs. The core average design discharge burnup for these bundles is 7000 MW·d/Te U and maximum burnup for assembly goes upto of 15 000 MWD/Te U. Use of fuel materials like MOX, Thorium, slightly enriched uranium etc in place of natural uranium in 19-element fuel bundles, in 220 MW(e) PHWRs is being investigated to achieve higher burnups. The maximum burnup investigated with these bundles is 30 000 MW·d/Te U. In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered. These studies for the PHWR fuel elements/assemblies have been elaborated in this paper. (author)

  3. The SINQ project. High flux without enrichment problems?

    International Nuclear Information System (INIS)

    A spallation neutron source (SINQ) designed for operation in a continuous mode is presently under construction at The Paul Scherrer Institute in Switzerland and is scheduled for completion in 1994. The waste beam from an isochronous proton cyclotron which is being upgraded to deliver around 1.5 mA of 590 MeV protons will be used after passing through two targets for meson production. The neutron flux in the D2O moderator surrounding the Pb-Bi spallation target is anticipated to amount to some 1014 n/cm2s per mA of beam current on target. This will put SINQ in the regime of most of the medium flux reactors which are presently being considered for operation with low fuel enrichment. Since SINQ will be the world's first spallation neutron source of that design and power level it is difficult to predict, how far the technology can be taken to build spallation neutron sources whose time average neutron flux would reach to level of advanced high flux reactors now possible with highly enriched uranium as fuel material. Another uncertainty factor is the reliability and operational stability of an accelerator needed to achieve such a high power source, which also has never been built so far. Taking everything together, it must be stated that at the very minimum a substantial development effort in a number of fields would be required, most likely leading over various intermediate stages before spallation neutron sources can compete in flux level with advanced high flux reactor designs. (orig.)

  4. Depletion modeling of integral burnable absorbers containing enriched boron

    International Nuclear Information System (INIS)

    Depletion modeling of fuel assembly with different integral fuel burnable absorber loadings containing enriched boron has been performed by WIMSD transport code. Equivalent boron concentration that represents depletion of the integral fuel burnable absorbers containing enriched boron has been calculated using modified PSU/LEOPARD code. The calculated equivalent boron concentrations have been introduced into FUMACS computer code package master files, upgrading the code package with new global calculation feature for core modeling with different integral fuel burnable absorber loadings containing enriched boron. This new feature of FUMACS/FEEC2001 code package has been verified and validated on 12-month and 18-month operating cycle core loading patterns of NPP Krsko.(author)

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

    International Nuclear Information System (INIS)

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

  6. Enriched uranium recovery at Los Alamos

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Brim, Cornelia P.

    2013-04-01

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

  10. The Whole-Core LEU U3Si2-Al Fuel Demonstration in the 30-MW Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U3Si2-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235U burnups, validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235U burnup support the corresponding measured quantities. In general, calculations for 60Co and 198Au reaction rate distributions, differential and integral control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 44 refs., 57 figs., 45 tabs

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

    International Nuclear Information System (INIS)

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

  12. Uranium thorium dioxide fuel-cycle and economic analysis

    International Nuclear Information System (INIS)

    The fuel division of Framatome ANP (Advanced Nuclear Power) is performing a fuel-cycle analysis for uranium-thorium dioxide (U/Th) reactor fuel as part of a U.S. Department of Energy Nuclear Energy Research Initiative project titled, ''Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactor'', (DE-FC03-99SF21916). The objective is to evaluate the economic viability of the U/Th fuel cycle in commercial nuclear reactors operating in the U.S. This analysis includes formulating the evaluation methodology, validating the methodology via benchmark calculations, and performing a fuel-cycle analysis and corresponding economic evaluation. The APOLLO2-F computer program of Framatome ANP SCIENCE package was modified to incorporate the thorium decay chains and provide cross sections for the SCIENCE fuel-cycle analysis. A comparison and economic evaluation was made between UO2 and UO2/ThO2 fuel cycles in a typical 193-fuel assembly pressurized water reactor using reload batch sizes corresponding to batch average discharge burnups of 50, 70, and 90 GWd/mtHM. Results show an increase in front-end costs for the UO2/ThO2 cycles due primarily to the higher cost in separative work units for enriching the uranium to 19.5 wt% 235U. (author)

  13. TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel

    International Nuclear Information System (INIS)

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

  14. How is uranium supply affecting enrichment?

    International Nuclear Information System (INIS)

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

  15. Enrichment: Dealing with overcapacity

    International Nuclear Information System (INIS)

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

  16. Laser and gas centrifuge enrichment

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-09

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

  17. Laser and gas centrifuge enrichment

    Science.gov (United States)

    Heinonen, Olli

    2014-05-01

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

  18. Evolution of fuel cycles for NPP with WWER-440. Status and Prospects

    International Nuclear Information System (INIS)

    In this paper main ways for WWER-440 fuel assemblies (FA) and fuel cycle improving are discussed and various examples are shown. Based on the presented results, the following conclusions have been made: 1) Operational reliability of this new fuel has the same level, as traditional fuel. 2) Technical solutions laid in the design of second-generation fuel assemblies, were proven and confirmed by the results of trial and commercial operation. 3) Development of a technical proposal on a third-generation fuel assembly is currently underway. 4) Profiled fuel with burnable absorber based on gadolinium makes it possible to realize full-scale 5-year fuel cycles with average fuel enrichment reduced from 4.4-4.38% to 4.25% (in second-generation assemblies). 5) Present-day fuel cycles for WWER-440 developed on the base of new FA constructions (second-generation fuel) ensure considerable increase of nuclear fuel utilization efficiency. 6) Present-day fuel cycles for WWER-440 make it possible to realize various operational fuel load lifetimes. This allows optimal adapting of the unit's electricity production to the specific energy system requirements. 7) Fuel cycles for WWER-440 using modernized construction of fuel were developed. These cycles make it possible to work at excess reactor power of up to 105-110%. 8) In case the modernized fuel is used, present-day fuel cycles for WWER-440 also make it possible to realize the maneuvering reactor operation mode, when the reactor power varies in wide frames during a short time period

  19. Advanced uranium enrichment processes

    International Nuclear Information System (INIS)

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

  20. Segal Enriched Categories I

    CERN Document Server

    Bacard, Hugo V

    2010-01-01

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

  1. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor can 'burn' a wide range of fuels without modification to the reactor system, including natural uranium, slightly enriched uranium, mixed oxide and spent LWR fuels. The economic feasibility of the advanced fuel cycles requires consideration of their disposal costs. Preliminary cost analyses for the disposal of spent CANDU-SEU (Slightly Enriched Uranium) and CANDU-DUPIC (Direct Use of spent PWR fuel In CANDU) fuels have been performed and compared to the internationally published costs for the direct disposal of spent CANDU and LWR fuels. The analyses show significant economic advantages in the disposal costs of CANDU-SEU and CANDU-DUPIC fuels. (author)

  2. Profile of World Uranium Enrichment Programs-2009

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    2009-04-01

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

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

    OpenAIRE

    Gruber, N.; Graven, H. D.

    2011-01-01

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

  4. Novovoronezh Unit 5 WWER-1000 reactor transfer to uranium-gadolinium fuel -operating experience of fuel cycles 23 and 24

    International Nuclear Information System (INIS)

    Development of Uranium-Gadolinium Fuel (UGF) cycle and safety justification for the Novovoronezh NPP Unit 5 were performed by Kurchatov Institute, Gidropress, VNIINM, VNIIAES, while preparation of the new FAs for fabrication was performed by NZKhK. The purpose of the project was to support fuel cycle duration of 300 effective days and longer with the average burn-up of about 50 MWday/kgU in spent FAs. The work was coordinated by concerns TVEL and Rosenergoatom. Special UGF FAs were developed for the Novovoronezh Unit 5 with initial enrichment of 3.9% and 4.3%. Core maps with UGF FAs are provided. During pilot operation (fuel loads 23 to 26) installation of UGF FAs was based on full scope of the core make-up. Fuel load 23 operated from 29 August 2005 to 25 July 2006. The fuel cycle duration was 312.5 effective days. The second batch of UGF FAs operated in fuel load 24 from 4 September 2006 to 14 July 2007. 30 UGF FAs were loaded in the core with the average initial enrichment of 4.3%, and 12 UGF FAs - with the average initial enrichment of 3.9%. This fuel cycle lasted for 302 effective days. For both cycles reactor power behavior curve and the maximum relative FA power as well as the curves of calculated and measured boric acid concentrations by chemical analysis during the cycles are shown. Comparison of FA relative power reconstructed by thermocouple (ThC) readings with calculation in maximum-density FA, reconstructed power root mean-square deviation (RMSD) from calculation in self-powered neutron detectors (SPND) locations during 23rd and 24rd fuel charge operation are presented. Changes in reactor plant thermal power and FA maximum relative power during the both fuel campaigns are also given. Based on the presented results authors concluded that: 1) During UGF operation all parameters monitored according to the Novovoronezh NPP Unit 5 technical specification of safe operation were within operational limits; 2) Introduction of UGF FAs did not result in increased

  5. Average Range and Network Synchronizability

    International Nuclear Information System (INIS)

    The influence of structural properties of a network on the network synchronizability is studied by introducing a new concept of average range of edges. For both small-world and scale-free networks, the effect of average range on the synchronizability of networks with bounded or unbounded synchronization regions is illustrated through numerical simulations. The relations between average range, range distribution, average distance, and maximum betweenness are also explored, revealing the effects of these factors on the network synchronizability of the small-world and scale-free networks, respectively. (general)

  6. Physical Theories with Average Symmetry

    CERN Document Server

    Alamino, Roberto C

    2013-01-01

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

  7. Enriching Number Knowledge

    Science.gov (United States)

    Mack, Nancy K.

    2011-01-01

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

  8. Everyone Vegetarian, World Enriching

    OpenAIRE

    Wu, John Y.

    2014-01-01

    This essay advocates global vegetarian diet. Firstly, seven personal health benefits and four global benefits of vegetarian dinners are specified to enrich the entire globe, and then, secondly, I explore concretely how to overcome internal hurdle and external hurdle, so as to effectively propagate vegetarian dinners throughout the world. Everyone wins, including animals!

  9. Methodology for content enrichment

    NARCIS (Netherlands)

    Nederbragt, H.; Heerlien, M.

    2010-01-01

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

  10. Designing job enrichment projects.

    Science.gov (United States)

    Clakeley, G L

    1988-01-01

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

  11. Job Enrichment in Extension.

    Science.gov (United States)

    Fourman, Louis S.; Jones, Jo

    1997-01-01

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

  12. Falling the fuel assembly in core mesh of reactor

    International Nuclear Information System (INIS)

    Accident reflecting drop of a fuel assembly (FA) in core mesh during the overload operations in the INP AS RUz research reactor is observed. Calculations and analysis of the accident situation were carried out for the reactor cores formed from fully high enriched IRT-3M type fuel (36% enrichment on '235U), the first mixed core consisting from 16 IRT-3M and 4 IRT-4M with low enriched fuel (19.7% enrichment on 235U), and the core fully formed from low enriched fuel. (authors)

  13. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

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

  14. 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.

  15. Preliminary Neutronics Analysis Of Fuel Pebble With Thorium Fuel Cycle

    International Nuclear Information System (INIS)

    A new fuel pebble was designed based on Thorium fuel cycle. 231Pa has been added into fuel pebble for obtaining the minimum reactivity swing. The results show that the new designed pebble fuel with 7.0 % 233U enrichment adding 3.2% 231Pa, the keff is to be controlled up to 65 GWd/t; the other design with 8.0 % 233U enrichment requires 3.9% 231Pa, the keff therefore is remain up to 80 GWd/t. About 95% of loaded 231Pa in fuel pebble is depleted after 120 GWd/t. The results imply that it is optimistic to design the fuel pebble with 233U, 231Pa and 232Th; but some effects such as fuel temperature effect, distribution of TRISO particle in pebble fuel, etc. are required to investigate. (author)

  16. 77 FR 823 - Guidance for Fuel Cycle Facility Change Processes

    Science.gov (United States)

    2012-01-06

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

  17. "Pricing Average Options on Commodities"

    OpenAIRE

    Kenichiro Shiraya; Akihiko Takahashi

    2010-01-01

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

  18. The effect of changing enrichments on core performance

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Status of RERTR fuel demonstrations

    International Nuclear Information System (INIS)

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

  1. Economic study of fuel scenarios for a reload

    International Nuclear Information System (INIS)

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

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H B

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

  5. Nuclear calculation for employing medium enrichment in reactors of Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    The fuel used for the research reactors of Japan Atomic Energy Research Institute (JAERI) is presently highly enriched uranium of 93%. However, the U.S. government (the supplier of fuel) is claiming to utilize low or medium enriched uranium from the viewpoint of resistivity to nuclear proliferation, and the availability of highly enriched uranium is becoming hard owing to the required procedure. This report is described on the results of nuclear calculation which is the basis of fuel design in the countermeasures to the reduction of enrichment. The basic conception in the reduction of enrichment is three-fold: to lower the latent potential of nuclear proliferation as far as possible, to hold the present reactor performance as far as possible, and to limit the reduction in the range which is not accompanied by the modification of reactor core construction and cooling system. This time, the increase of the density and thickness of fuel plates and the effect of enrichment change to 45% on reactivity and neutron flux were investigated. The fuel of UAl sub(x) - Al system was assumed, which was produced by powder metallurgical method. The results of investigations on JRR-2 and JMTR reactors revealed that 45% enriched fuel does not affect the performances much. However, deterioration of the performances is not neglegible if further reduction is needed. In future, the influence of the burn-up effect of fuel on the life of reactor cores must be investigated. (Wakatsuki, Y.)

  6. Uranium enrichment management review. Final report

    International Nuclear Information System (INIS)

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

  7. Enrichment Strategies in Phosphoproteomics.

    Science.gov (United States)

    Leitner, Alexander

    2016-01-01

    The comprehensive study of the phosphoproteome is heavily dependent on appropriate enrichment strategies that are most often, but not exclusively, carried out on the peptide level. In this chapter, I give an overview of the most widely used techniques. In addition to dedicated antibodies, phosphopeptides are enriched by their selective interaction with metals in the form of chelated metal ions or metal oxides. The negative charge of the phosphate group is also exploited in a variety of chromatographic fractionation methods that include different types of ion exchange chromatography, hydrophilic interaction chromatography (HILIC), and electrostatic repulsion HILIC (ERLIC) chromatography. Selected examples from the literature will demonstrate how a combination of these techniques with current high-performance mass spectrometry enables the identification of thousands of phosphorylation sites from various sample types. PMID:26584921

  8. The influence of fuel assembly characteristics on reactor safety

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Yuesheng Fan

    2010-12-01

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

  10. What is Enrichment For (?)

    OpenAIRE

    Matsangos, P.

    2011-01-01

    It has long been held that schooling, the curriculum (hidden or otherwise), and the pedagogical practices pertaining to them are the perpetrators of social and economic inequalities in Western society. My interest surrounds the role of ‘enrichment’, as an auxiliary of education, in perpetuating or restricting social justice. As the enrichment coordinator within an inner city Sixth Form college, it became part of my remit to research the dispositions held by our largely underprivileged student...

  11. A feasibility study concerning the conversion of the TR-2 reactor from using highly enriched uranium to light enriched uranium

    International Nuclear Information System (INIS)

    A study has been made of the feasibility of converting the 5-MW TR-2 reactor at CNAEM to use fuel with uranium enrichment of 3O8-Al fuel meat with a uranium density in the range 2.3 to 3.0 g/cm3 in the fuel meat with meat thickness varying between 0.9 and 1.00 mm, the number of plates in the LEU element being reduced from 23 in the HEU element to 19 to 20 to maintain adequate cooling. Fuels within this density range are expected to be commercially available within the next two years. From the results of the study it appears to be feasible to safely operate the TR-2 reactor using LEU fuel without increased fuel cycle costs or decreased performance using U2O8 fuels with densities in the 2.3 to 3.0 gU/cm3 range. (author)

  12. Uranium enrichment technologies

    International Nuclear Information System (INIS)

    Several technically mature and commercially proven methods of separation are currently available on the enrichment market, such as the diffusion and centrifuge methods. According to the information available, the diffusion technologies, which have dominated the scene so far, will lose much of their influence in favor of the centrifuge technologies which, because of their special process characteristics, such as lower energy consumption and smaller economic plant sizes, can adapt themselves to the conditions of the market in a flexible and economic way. Other methods, such as the gas dynamics Helikon and separation nozzle techniques, and chemical enrichment are on the verge of commercial maturity. However, their economic performance and chances in the enrichment market cannot yet be assessed with any degree of precision. The prospects and possibilities of the laser method, which has received much backing recently, especially in the USA, are difficult to weigh in the light of the knowledge available to date. However, it is hardly to be expected that technically mature and commercially viable plants operating by this technology are going to be available before the turn of the century. (orig.)

  13. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    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)

  14. Power convergence of Abel averages

    OpenAIRE

    Kozitsky, Yuri; Shoikhet, David; Zemanek, Jaroslav

    2012-01-01

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

  15. High-average-power lasers

    International Nuclear Information System (INIS)

    The goals of the High-Average-Power Laser Program at LLNL are to develop a broad technology base for solid state lasers and to demonstrate high-average-power laser operation with more efficiency and higher beam quality than has been possible with current technology. Major activities are the zig-zag laser testbed and the gas-cooled-slab laser test bed. This section describes these activities as well as discussion of material development; nonlinear optics; laser materials, and applications

  16. Data feature: Fuel procurement

    International Nuclear Information System (INIS)

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

  17. FUEL ASSAY REACTOR

    Science.gov (United States)

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

    1962-12-25

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

  18. Enrichment marketplace - today (and tomorrow)

    International Nuclear Information System (INIS)

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

  19. US developments in technology for uranium enrichment

    International Nuclear Information System (INIS)

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

  20. Engineering studies on chemical uranium enrichment

    International Nuclear Information System (INIS)

    In multi-stage separation processes such as the Chemical Enrichment Process, back mixing phenomena are a matter of vital importance for understanding separation efficiency. Back mixing is composed of two parts, microscopic and macroscopic. The microscopic back mixing is known through molecular theory. The macroscopic back mixing effects are quantitatively determined using flow patterns obtained through computational calculations of the fluid dynamics, as well as experimental data. Judging from physical theory and experimental results, the sum of microscopic and the macroscopic back mixing terms is approximately 300μm in the Chemical Enrichment Process. This technology will also contribute to nuclear science in that it can be applied not only to the isotope separation of other elements and to other multi-stage separation processes, but also to the reprocessing of spent fuel. (author)