Development of quality assurance methods for low enriched fuel assemblies

International Nuclear Information System (INIS)

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

2010-01-01

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

The low-enrichment fuel development program

International Nuclear Information System (INIS)

Stahl, D.

1993-01-01

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

Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

International Nuclear Information System (INIS)

Giorsetti, D.R.

1987-01-01

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

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.

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1993-07-01

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

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.

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.

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1988-05-01

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

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

International Nuclear Information System (INIS)

Travelli, A.

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-08-01

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

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

International Nuclear Information System (INIS)

1983-08-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

The current state of the Russian reduced enrichment research reactors program

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

Fuel enrichment reduction for heavy water moderated research reactors

International Nuclear Information System (INIS)

McCulloch, D.B.

1984-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

Roglans-Ribas, J.

2008-01-01

Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

Axial blanket enrichment optimization of the NPP Krsko fuel

International Nuclear Information System (INIS)

Kromar, M.; Kurincic, B.

2001-01-01

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

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

International Nuclear Information System (INIS)

Travelli, A.

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Qualification status of LEU [low enriched uranium] fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.

1987-01-01

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

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

International Nuclear Information System (INIS)

Primm, Trent; Guida, Tracey

2010-01-01

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

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

International Nuclear Information System (INIS)

Rydell, R.J.

1990-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Status of the reduced enrichment for research reactors program in Argentina

International Nuclear Information System (INIS)

Perez, E.; Kohut, C.

2004-01-01

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

Low-enriched fuel particle performance review

International Nuclear Information System (INIS)

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

1978-08-01

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

Low-enriched fuel particle performance review

International Nuclear Information System (INIS)

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

1978-08-01

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

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

International Nuclear Information System (INIS)

Travelli, A.

1987-01-01

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

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

International Nuclear Information System (INIS)

Hickey, B.

1983-01-01

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

Slightly enriched uranium fuel for a PHWR

International Nuclear Information System (INIS)

Notari, C.; Marajofsky, A.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

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

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

International Nuclear Information System (INIS)

1993-07-01

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

Optimal pin enrichment distributions in nuclear reactor fuel bundles

International Nuclear Information System (INIS)

Lim, E.Y.

1976-01-01

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

Conversion of research reactors to low-enrichment uranium fuels

International Nuclear Information System (INIS)

Muranaka, R.G.

1983-01-01

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

Yalina booster subcritical assembly performance with low enriched uranium fuel

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry

2011-01-01

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

Yalina booster subcritical assembly performance with low enriched uranium fuel

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-07-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Refueling the RPI reactor facility with low-enrichment fuel

International Nuclear Information System (INIS)

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

1985-01-01

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

The use of medium enriched uranium fuel for research reactors

International Nuclear Information System (INIS)

1979-01-01

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

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)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

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

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

International Nuclear Information System (INIS)

1993-07-01

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

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

International Nuclear Information System (INIS)

Vittorio, D.; Durance, G.

2002-01-01

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

Reduced enrichment program for the FRM-II, status 2004

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

Carter, Robert E.; Leonard, Bobby E.

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-08-01

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

Determining method and device for enrichment distribution inside of fuel assembly

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Hennebach, M.; Schnorrenberg, N.

2008-01-01

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

Progress in qualifying low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

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

2001-01-01

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

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.

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

International Nuclear Information System (INIS)

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

2006-03-01

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

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

OpenAIRE

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

2002-01-01

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

Status of reduced enrichment program for research reactors in Japan

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Determination of enrichment of recycle uranium fuels for different burnup values

International Nuclear Information System (INIS)

Zabunoglu, Okan H.

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-02-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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.

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

International Nuclear Information System (INIS)

de la Garza, A.

1977-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Science.gov (United States)

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

2018-06-01

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

The development of lower enrichment fuels for Canadian research reactors

Energy Technology Data Exchange (ETDEWEB)

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

1983-08-01

As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1993-08-01

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

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1993-08-01

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

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

International Nuclear Information System (INIS)

Heckendorn, F.M.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Experience with a fuel rod enrichment scanner

International Nuclear Information System (INIS)

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

1975-01-01

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

Analysis of a PHWR slightly enriched fuel

International Nuclear Information System (INIS)

Notari, C.; Marajofsky, A.

1994-01-01

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

Development for analysis system of rods enrichment of nuclear fuels

International Nuclear Information System (INIS)

Rojas C, E.L.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO 2 required. (Author)

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

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

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

International Nuclear Information System (INIS)

Liu Lanhua; Liu Nangai

1990-07-01

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

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

International Nuclear Information System (INIS)

Kelly, J.L.

1978-09-01

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

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

International Nuclear Information System (INIS)

Rothwell, Geoffrey

2008-01-01

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

Reduced enriched fuel status at CERCA

International Nuclear Information System (INIS)

Tissier, A.; Fanjas, Y.

1991-01-01

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

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

International Nuclear Information System (INIS)

1978-10-01

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

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

Directory of Open Access Journals (Sweden)

Mohammed Saad Dwiddar

2015-08-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

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

International Nuclear Information System (INIS)

Caprioli, Sara

2004-04-01

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

Configuration of LWR fuel enrichment or burnup yielding maximum power

International Nuclear Information System (INIS)

Bartosek, V.; Zalesky, K.

1976-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1987-10-01

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

Advances in uranium enrichment processes

International Nuclear Information System (INIS)

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

1986-05-01

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

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

International Nuclear Information System (INIS)

Saragih, Tukiran; Pinem, Surian

2002-01-01

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

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

International Nuclear Information System (INIS)

Travelli, A.

1986-11-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Low enrichment fuel conversion for Iowa State University. Final report

International Nuclear Information System (INIS)

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

1996-01-01

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

Low-enriched uranium-molybdenum fuel plate development

International Nuclear Information System (INIS)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the technical specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort. (author)

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

International Nuclear Information System (INIS)

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

1991-01-01

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort

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

International Nuclear Information System (INIS)

Broadhead, B.L.

1998-08-01

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

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

International Nuclear Information System (INIS)

Vitkova, Marietta; Gorinov, Ivan

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

International Nuclear Information System (INIS)

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

1985-06-01

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

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

International Nuclear Information System (INIS)

Hassel, H.W.

1993-01-01

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

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.

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

International Nuclear Information System (INIS)

Sato, S.

1975-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Development of IAEA safeguards at low enrichment uranium fuel fabrication plants

International Nuclear Information System (INIS)

Badawy, I.

1988-01-01

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

Status of reduced enrichment programs for research reactors in Japan

International Nuclear Information System (INIS)

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

1997-01-01

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

Status of reduced enrichment programs for research reactors in Japan

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

1980-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Minimization of waste from uranium purification, enrichment and fuel fabrication

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-10-01

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

Minimization of waste from uranium purification, enrichment and fuel fabrication

International Nuclear Information System (INIS)

1999-10-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

1978-09-01

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

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

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2013-10-24

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

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

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2013-04-18

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

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

International Nuclear Information System (INIS)

Moss, T.A.

1993-01-01

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

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

Science.gov (United States)

Bhattacharya, Sumit

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

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

International Nuclear Information System (INIS)

Estryk, G.; Higa, M.

1993-01-01

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

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

International Nuclear Information System (INIS)

Schwartz, J.P.

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

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

International Nuclear Information System (INIS)

Pazirandeh, A.; Nasiri, S. H.

2012-01-01

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

The burnable poisons utilization for fissile enriched CANDU fuel bundle

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1984-05-01

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

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

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1984-05-01

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

Uranium enrichment

International Nuclear Information System (INIS)

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

1988-06-01

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

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

International Nuclear Information System (INIS)

Hida, Kazuki; Yoshioka, Ritsuo

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1987-02-01

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

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

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

Science.gov (United States)

2012-03-27

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

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

Science.gov (United States)

2012-10-30

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

Fuel assembly

International Nuclear Information System (INIS)

Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

1998-01-01

In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

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

International Nuclear Information System (INIS)

Deng Jingshan; Zhou Chengfang; Luo Minxuan; Liu Yun

1997-01-01

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

Fuel assembly

International Nuclear Information System (INIS)

Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

1997-01-01

In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

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

International Nuclear Information System (INIS)

Yapici, H.; Baltacioglu, E.

1997-01-01

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

Draining Water from Aircraft Fuel Using Nitrogen Enriched Air

Directory of Open Access Journals (Sweden)

Michael Frank

2018-04-01

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

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

International Nuclear Information System (INIS)

Mitchell, M.E.

1976-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

Raica, V.; Sindile, A.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

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

International Nuclear Information System (INIS)

Roh, Seungkook; Kim, Wonjoon

2014-01-01

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

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

International Nuclear Information System (INIS)

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

2010-01-01

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

Oxygen enrichment incineration

International Nuclear Information System (INIS)

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

2000-10-01

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

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.

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

International Nuclear Information System (INIS)

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

1991-07-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Argentine activities related to the development of low enriched fuel elements

International Nuclear Information System (INIS)

Giorsetti, Domingo R.; Perez, Edmundo E.

1983-01-01

Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the U.S.A. and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project), has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown in Table 1. At a meeting held in Vienna in March, 1980, the CNEA stated that its development of fuels with low enrichment would be in two fuel lines: UAl x -Al and U 3 O 8 -Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm 3 for the UAI x -Al line and 2.4-3.0 g/cm 3 for the U 3 O 8 line. At the international meeting held at ANL in November, 1980, and after having received depleted uranium and uranium with 20% and 45% enrichment (purchased from the U.S.A. for manufacturing miniplates and possible standard fuels) to carry on the proposed development, CNEA anticipated -- after its first tests -- that the conditions were satisfactory for reaching uranium densities of 2.4-3.0 g/cm 3 in U 3 O 8 -Al fuel and of 2.4 g/cm 3 in UAI x -Al fuel. In February 1981, after Argentina accepted the obligation of paying for the irradiation service, authorization was obtained for irradiating miniplates in the Oak Ridge Reactor within the RERTR Program. In June 1981, the first set of miniplates was sent to Oak Ridge National Laboratory (ORNL). The maximum actual densities reached at that time were 3.12 g/cm 3 with U 3 O 8 -Al and 2.52 g/cm 3 with UAl x -Al. During a visit of the CNEA Project Technical Manager to the Argonne National Laboratory (ANL) in July 1981, and after exchanging ideas with ANL professional staff, the CNEA decided to incorporate a new line of development, that of U 3 Si-Al. Three months later

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

International Nuclear Information System (INIS)

Hassel, H.W.

1993-01-01

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

US enrichment reduction studies

International Nuclear Information System (INIS)

1979-06-01

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

IAEA activities related to research reactor fuel conversion and spent fuel return programs

International Nuclear Information System (INIS)

Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

2005-01-01

The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

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

International Nuclear Information System (INIS)

Zuhair; Suwoto

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1979-04-01

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

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

International Nuclear Information System (INIS)

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

1979-04-01

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

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.

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.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Solid oxide fuel cells fueled with reducible oxides

Science.gov (United States)

Chuang, Steven S.; Fan, Liang Shih

2018-01-09

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Development of ISA procedure for uranium fuel fabrication and enrichment facilities

International Nuclear Information System (INIS)

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

2011-01-01

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

The low enriched uranium fuel cycle in Ontario

International Nuclear Information System (INIS)

Archinoff, G.H.

1979-02-01

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

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

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

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

International Nuclear Information System (INIS)

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

1985-05-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1991-10-01

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

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.

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

James, R.A.

1980-01-01

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

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

Science.gov (United States)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Chambers, Alex; Ragusa, Jean C.

2014-01-01

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

Use of enriched uranium as a fuel in CANDU reactors

International Nuclear Information System (INIS)

Zech, H.J.

1976-08-01

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

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

International Nuclear Information System (INIS)

1997-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-08-01

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

Fabrication of fuel elements on the basis of increased concentration fuel composition

International Nuclear Information System (INIS)

Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

2004-01-01

As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

IAEA activities related to research reactor fuel conversion and spent fuel return programmes

International Nuclear Information System (INIS)

Ritchie, I.G.; Adelfang, P.; Goldman, I.N.

2004-01-01

Full text: The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country of origin where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programmes have supported research reactor fuel conversion from HEU to low enriched uranium, and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. It is hoped that an announcement of the extension of the U.S. Acceptance Programme, which is expected in the very near future, will facilitate the life extensions of many productive TRIGA reactors around the world. (author)

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.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Development of nuclear fuel cycle technology

International Nuclear Information System (INIS)

Kawahara, Akira; Sugimoto, Yoshikazu; Shibata, Satoshi; Ikeda, Takashi; Suzuki, Kazumichi; Miki, Atsushi.

1990-01-01

In order to establish the stable supply of nuclear fuel as an important energy source, Hitachi ltd. has advanced the technical development aiming at the heightening of reliability, the increase of capacity, upgrading and the heightening of performance of the facilities related to nuclear fuel cycle. As for fuel reprocessing, Japan Nuclear Fuel Service Ltd. is promoting the construction of a commercial fuel reprocessing plant which is the first in Japan. The verification of the process performance, the ensuring of high reliability accompanying large capacity and the technical development for recovering effective resources from spent fuel are advanced. Moreover, as for uranium enrichment, Laser Enrichment Technology Research Association was founded mainly by electric power companies, and the development of the next generation enrichment technology using laser is promoted. The development of spent fuel reprocessing technology, the development of the basic technology of atomic process laser enrichment and so on are reported. In addition to the above technologies recently developed by Hitachi Ltd., the technology of reducing harm and solidification of radioactive wastes, the molecular process laser enrichment and others are developed. (K.I.)

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

International Nuclear Information System (INIS)

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

1984-02-01

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

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

International Nuclear Information System (INIS)

Nochev, T.

1999-01-01

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

Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

International Nuclear Information System (INIS)

Simanullang, Irwan Liapto; Obara, Toru

2017-01-01

Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-06-15

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

Advanced Neutron Source enrichment study

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Weeks, J.R.

1996-01-01

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

Fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto; Ogiya, Shunsuke.

1989-01-01

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

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

International Nuclear Information System (INIS)

Wu, Haicheng; Okumura, Keisuke; Shibata, Keiichi

2005-06-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Uranium Enrichment, an overview

International Nuclear Information System (INIS)

Coates, J.H.

1994-01-01

This general presentation on uranium enrichment will be followed by lectures on more specific topics including descriptions of enrichment processes and assessments of the prevailing commercial and industrial situations. I shall therefore avoid as much as possible duplications with these other lectures, and rather dwell on: some theoretical aspects of enrichment in general, underlying the differences between statistical and selective processes, a review and comparison between enrichment processes, remarks of general order regarding applications, the proliferation potential of enrichment. It is noteworthy that enrichment: may occur twice in the LWR fuel cycle: first by enriching natural uranium, second by reenriching uranium recovered from reprocessing, must meet LWR requirements, and in particular higher assays required by high burn up fuel elements, bears on the structure of the entire front part of the fuel cycle, namely in the conversion/reconversion steps only involving UF 6 for the moment. (author). tabs., figs., 4 refs

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

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-04-01

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

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

International Nuclear Information System (INIS)

Pesic, M. P.

1997-01-01

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

Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

2009-01-01

A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

Use of highly enriched uranium in the material testing reactor BR2

International Nuclear Information System (INIS)

Beeckmans de West-Meerbeeck, A.

1979-05-01

In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided; if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly

Use of highly enriched uranium in the material testing reactor BR2

International Nuclear Information System (INIS)

Beeckmans de West-Meerbeeck, A.

1979-05-01

In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided: if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly. (author)

The Caramel fuel in OSIRIS

International Nuclear Information System (INIS)

Cherruau, Francois.

1980-11-01

This paper presents the main characteristics of the caramel fuel, a description of OSIRIS transformations that were decided in line with its conversion and the results of its operation since then. The Caramel fuel is made from sintered UO 2 pellets contained in zircaloy clads forming the plates of the fuel assembly reducing the enrichment need to as little as 3 to 10% instead of 93% enriched U/Al in the previous fuel. The first year of experience shows the capacity under a statistic scale of the caramel fuel to fulfil the most severe operation requirements for use in low and medium power research reactors

Characteristics of several equilibrium fuel cycles of PWR

International Nuclear Information System (INIS)

Waris, Abdul; Sekimoto, Hiroshi

2001-01-01

This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium. (author)

From high to low. The IAEA is helping to reduce the use of high-risk nuclear fuel at the world's research reactors

International Nuclear Information System (INIS)

Adelfang, P.; Goldman, I.

2006-01-01

Research reactors play a key role in the development of peaceful uses of atomic energy. They are used for the production of isotopes for medicine and industry, for research in physics, biology and materials science, and for scientific education and training. They also continue to play an important role in support of nuclear power programmes. The IAEA's data shows there are 249 operational research reactors worldwide. Of these, more than 100 reactors are still fuelled with highly enriched uranium (HEU). It is considered high-risk nuclear material since it can be easily used for a nuclear explosive device. As part of a developing international norm to minimize and eventually eliminate HEU in civilian nuclear applications, research reactor operators increasingly are working with national and international agencies. They are being encouraged and supported to improve their physical security arrangements, convert their reactors to low-enriched uranium (LEU) fuel, and ship irradiated fuel back to the country of origin.For more than twenty years the IAEA has been supporting international efforts associated with reducing the amount of HEU in international commerce. Projects and activities have directly supported a programme the United States initiated in 1978, called Reduced Enrichment for Research and Test Reactors (RERTR). The IAEA's work additionally supports efforts to return research reactor fuel to the country where it was originally enriched so-called take back activities. IAEA initiatives have included the development and maintenance of several databases with information related to research reactors and research reactor spent fuel inventories. These databases have been essential in planning and managing both RERTR and take-back programmes. Other Agency activities through technical cooperation and other channels have supported the conversion of research reactors to using lower enriched fuels. In other ways, the IAEA supports the exchange of information among experts

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

International Nuclear Information System (INIS)

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

2009-01-01

A concept of 'minimal-content gadolinia' with a content of less than several hundred ppm mixed in the 'above-5wt% enrichment UO 2 fuel' for super high burnup is proposed for ensuring the criticality safety in the UO 2 fuel fabrication facility for light water reactors (LWRs) without increase in investment cost. Required gadolinia contents calculated were from 53 to 305 ppm for enrichments of UO 2 powders for boiling water reactor (BWR) fuel from 6 to 10 wt%. It is expected that the minimal-content gadolinia yields an acceptable reactivity suppression at the beginning of operating cycle and no reactivity penalty at the end of operating cycle due to no residual gadolinium. A series of critical experiments were carried out in the Toshiba Nuclear Critical Assembly (NCA). Reactivity effects of the gadolinia were measured to clarify the nuclear characteristics, and the measured values and the calculated values agreed within 5%. (author)

Volatile behaviour of enrichment uranium in the total nuclear fuel price

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Persiani, P.J.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1978-08-01

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

Multi-stage combustion using nitrogen-enriched air

Science.gov (United States)

Fischer, Larry E.; Anderson, Brian L.

2004-09-14

Multi-stage combustion technology combined with nitrogen-enriched air technology for controlling the combustion temperature and products to extend the maintenance and lifetime cycles of materials in contact with combustion products and to reduce pollutants while maintaining relatively high combustion and thermal cycle efficiencies. The first stage of combustion operates fuel rich where most of the heat of combustion is released by burning it with nitrogen-enriched air. Part of the energy in the combustion gases is used to perform work or to provide heat. The cooled combustion gases are reheated by additional stages of combustion until the last stage is at or near stoichiometric conditions. Additional energy is extracted from each stage to result in relatively high thermal cycle efficiency. The air is enriched with nitrogen using air separation technologies such as diffusion, permeable membrane, absorption, and cryogenics. The combustion method is applicable to many types of combustion equipment, including: boilers, burners, turbines, internal combustion engines, and many types of fuel including hydrogen and carbon-based fuels including methane and coal.

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

KAUST Repository

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

2012-01-01

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

PWR-to-PWR fuel cycle model using dry process

International Nuclear Information System (INIS)

Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

2002-03-01

PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

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

International Nuclear Information System (INIS)

Serghiuta, D.

1994-01-01

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

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

International Nuclear Information System (INIS)

1986-06-01

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

Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

Energy Technology Data Exchange (ETDEWEB)

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

1985-07-01

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

Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

International Nuclear Information System (INIS)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

Coyne, P.; Smith, G.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Uranium enrichment

International Nuclear Information System (INIS)

Mohrhauer, H.

1982-01-01

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

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

Energy Technology Data Exchange (ETDEWEB)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N. [Westinghouse Electric Co. LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

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

International Nuclear Information System (INIS)

Jacquemin, J.

1985-01-01

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

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

International Nuclear Information System (INIS)

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

1987-10-01

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

Towards proliferation-resistant thorium fuels

International Nuclear Information System (INIS)

Alhaj, M. Yousif; Mohamed, Nader M.A.; Badawi, Alya; Abou-Gabal, Hanaa H.

2017-01-01

Thorium-plutonium mixture is proposed as alternative nuclear reactor fuel to incinerate the increasing stockpile plutonium. However, this fuel will produce an amount of uranium with about 90% 233U at applicable discharge burnups (60GWD/MTU). This research focuses on proposing an optimum non proliferative thorium fuel, by adding a small amount of 238U to reduce the attractiveness of the resultant uranium. Three types of additive which contain 238U were used: 4.98% enriched, natural and depleted uranium. We found that introducing uranium to the fresh thorium-plutonium fuel reduces its performance even if the uranium was enriched up to 5%. While uranium admixtures reduce the quality of the reprocessed uranium, it also increases the quality of the plutonium. However, this increase is very low compared to the reduced quality of uranium. We also found that using uranium as admixture for thorium-plutonium mixed fuel increases the critical mass of the extracted uranium by a factor of two when using only 1% admixture of uranium. The higher the percentage of uranium admixture the higher the critical mass of the reprocessed one.

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

Carter, R.E.

1985-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Carter, R E

1985-07-01

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

Fuel assembly and nuclear reactor core

International Nuclear Information System (INIS)

Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

1995-01-01

The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

International Nuclear Information System (INIS)

Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

2009-01-01

The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

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

International Nuclear Information System (INIS)

Guelol Oezdere, Oya

2003-01-01

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)

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)

Future of uranium enrichment

International Nuclear Information System (INIS)

Hosmer, C.

1981-01-01

The increasing amount of separative work being done in government facilities to produce low-enriched uranium fuel for nuclear utilities again raises the question: should this business-type, industrial function be burned over the private industry. The idea is being looked at by the Reagan administration, but faces problems of national security as well as from the unique nature of the business. This article suggests that a joint government-private venture combining enriching, reprocessing, and waste disposal could be the answer. Further, a separate entity using advanced laser technology to deplete existing uranium tails and lease them for fertile blankets in breeder reactors might earn substantial revenues to help reduce the national debt

Uranium enrichment

International Nuclear Information System (INIS)

1990-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

Misra, B.; Davis, J.F.

1980-02-01

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

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

International Nuclear Information System (INIS)

Muellauer, J.

1989-01-01

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

A study of UO2 wafer fuel for very high-power research reactors

International Nuclear Information System (INIS)

Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

1983-01-01

The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-09-30

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

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

Science.gov (United States)

Liu, Hongyan; Wang, Hongyu

2016-07-01

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

Theoretical studies aiming at the IEA-R1 reactor core conversion from high U-235 enrichment to low U-235 enrichment

International Nuclear Information System (INIS)

Frajndlich, R.

1982-01-01

The research reactors, of which the fuel elements are of MTR type, functions presently, almost in their majority with high U-235 enrichment. The fear that those fuel elements might generate a considerabLe proliferation of nuclear weapons rendered almost mandatory the conversion of highly enriched fuel elements to a low U-235 enrichment. As the IEA-R1 reactor of IPEN is operating with highly enriched fuel elements a study aiming at this conversion was done. The problems related to the conversion and the results obtained, demonstrated the technical viabilty for its realization. (E.G.) [pt

Promotion of uranium enrichment business

International Nuclear Information System (INIS)

Kurushima, Morihiro

1981-01-01

The Committee on Nuclear Power has studied on the basic nuclear power policy, establishing its five subcommittees, entrusted by the Ministry of Nternational Trade and Industry. The results of examination by the subcommittee on uranium enrichment business are given along with a report in this connection by the Committee. In order to establish the nuclear fuel cycle, the aspect of uranium enrichment is essential. The uranium enrichment by centrifugal process has proceeded steadily in Power Reactor and Nuclear Fuel Development Corporation. The following matters are described: the need for domestic uranium enrichment, the outlook for overseas enrichment services and the schedule for establishing domestic enrichment business, the current state of technology development, the position of the prototype enrichment plant, the course to be taken to establish enrichment business the main organization operating the prototype and commercial plants, the system of supplying centrifuges, the domestic conversion of natural uranium the subsidies for uranium enrichment business. (J.P.N.)

IFBA credit in the Shearon Harris fuel racks with Vantage 5 fuel

International Nuclear Information System (INIS)

Boyd, W.A.; Schmidt, R.F.; Erwin, R.D.

1989-01-01

At the Shearon Harris nuclear plant, fuel management strategies are being considered which will result in feed fuel enrichments approaching 5.0 w/o U-235. These types of enrichments require a new criticality analysis to raise the existing fuel rack enrichment limit. It is receiving Westinghouse Vantage 5 fuel with integral fuel burnable absorber (IFBA) rods providing the depletable neutron absorber. An analysis was performed on the fuel racks which demonstrates that fuel enriched up to 5.0 w/o U-235 can be stored by taking credit for the IFBA rods present in the high enriched fuel assemblies. This is done by calculating the maximum Vantage 5 fuel assembly reactivity that can be placed in the fuel racks and meet the criticality K-eff limit. A methodology is also developed which conservatively calculates the minimum number of IFBA rods needed per assembly to meet the fuel rack storage limits. This eliminates the need for core designers to determine assembly K-inf terms for every different enrichment/IFBA combination

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

International Nuclear Information System (INIS)

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

1988-07-01

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

Comparison of DUPIC fuel composition heterogeneity control methods

International Nuclear Information System (INIS)

Choi, Hang Bok; Ko, Won Il

1999-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-02-15

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

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

International Nuclear Information System (INIS)

Dobrevski, I.; Zaharieva, N.

2011-01-01

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

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

Science.gov (United States)

2016-07-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Experimental RA reactor operation with 80% enriched fuel - Program of experimental operation: a) Program of experimental operation with 80% enriched fuel at low power, b) contents of the experimental operation with 80% enriched fuel at higher power levels; Program probnog rada: a) Program probnog rada reaktora sa 80% obogacenim gorivom na malim snagama, b) sadrzaj programa probnog rada reaktora RA sa 80% obogacenim gorivom na vecim snagama

Energy Technology Data Exchange (ETDEWEB)

Martinc, R; Sotic, O; Skoric, M; Cupac, S; Bulovic, V; Maric, I; Marinkov, L [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1980-10-15

Highly enriched (80%) uranium oxide fuel was regularly used in the mixed reactor core with the 2% enriched fuel since 1976. The most important changes related to reactor operation, in comparison with the original design project were related to reactor core fuelling schemes. At the end of 1979 reactor was shutdown due to the corrosion coating noticed on some fuel elements and due to decrease quality of the heavy water. Subsequently the Sanitary inspector of Serbia has prohibited further reactor operation. Restart of the reactor will not be a simple continuation of operation. It is indispensable to perform complete experimental program including measurements of critical parameters at different power levels for the core with fresh 80% enriched fuel. The aim of this document is to obtain working permission and its contents are in agreement with the procedure demanded by the Safety Committee of the Institute. It includes results of optimization and safety analysis for the initial reactor core. Since the permission for restart is not obtained, a separate RA reactor safety report is prepared in addition to the program for experimental operation. This report includes: detailed program for reactor experimental operation with 80% enriched fuel in the core at low power levels, and contents of the experimental operation with 80% enriched fuel in the core at higher power levels. [Serbo-Croat] Od decembra 1976. godine redovno je korisceno 80% obogaceno gorivo u mesanoj resetki reaktorskog jezgra sa 2% obogacenim gorivom. Najvece izmene na reaktoru u odnosu na originalni projekat izvrsene su u nacinu rukovanja gorivom. Krajem marta 1979. godine obustavljen je rad reaktora usled naslaga na gorivnim elementima i loseg stanja teske vode. Naknadno je izdata zabrana za rad reaktora od strane Sanitarnog inspektora SR Srbije. Ponovno pustanje reaktora u rad nece biti jednostavan nastavak rada. Neophodno je da se izvede kompletan program merenja kriticnih parametara i drugih

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

Directory of Open Access Journals (Sweden)

Hammam Oktajianto

2014-12-01

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

Analysis of neutronic parameters related to reduction in fuel rod diameter for Angra-1 reactor fuel elements

International Nuclear Information System (INIS)

Faria, Eduardo F.; Sadde, Luciano M.; Sakai, Massao; Gomes, Sydney da S.

2000-01-01

The actual fuel element design for Angra-1 PWR satisfies in a very conservative way the design limits established for the critical heat flux as well as for the energy stored in the fuel rod. However, that is not an optimized design under neutronic considerations. The conservative ratio of the H and U atomic densities gives rise to a harder neutron spectrum which reduces its reactivity. In this report, a reduction in fuel rod diameters has been analyzed, keeping however the same rod pitch for geometrical compatibility reasons. By increasing the H/U ratio it is possible to obtain a net gain in reactivity. The optimized diameter in its turn should not jeopardize the reactor safety requirements. The actual trends of the nuclear industry is to extend the cycles and the enrichment by using advanced fuel design. It must be emphasized that this design change gives rise to economical advantages, for example, reduced costs for uranium utilization and enrichment with a net gain in reactivity. (author)

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

International Nuclear Information System (INIS)

1978-01-01

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

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)

Blueprint for domestic uranium enrichment

International Nuclear Information System (INIS)

1981-01-01

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.)

Results of fuel elements fabrication on the basis of increased concentration dioxide fuel for research reactors

International Nuclear Information System (INIS)

Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

1996-01-01

According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)

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

International Nuclear Information System (INIS)

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

1984-01-01

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

PC based uranium enrichment analyser

International Nuclear Information System (INIS)

Madan, V.K.; Gopalakrishana, K.R.; Bairi, B.R.

1991-01-01

It is important to measure enrichment of unirradiated nuclear fuel elements during production as a quality control measure. An IBM PC based system has recently been tested for enrichment measurements for Nuclear Fuel Complex (NFC), Hyderabad. As required by NFC, the system has ease of calibration. It is easy to switch the system from measuring enrichment of fuel elements to pellets and also automatically store the data and the results. The system uses an IBM PC plug in card to acquire data. The card incorporates programmable interval timers (8253-5). The counter/timer devices are executed by I/O mapped I/O's. A novel algorithm has been incorporated to make the system more reliable. The application software has been written in BASIC. (author). 9 refs., 1 fig

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Mishima, Kaichiro; Kanda, Keiji; Shibata, Toshikazu

1985-01-01

A feasibility study has been performed on the core conversion to the use of low-enriched uranium (LEU) fuels in the KUR. Five fuel element geometries are studied. For each fuel element, the relation between the pressure drop and the flow rate, critical heat flux, and heat fluxes for the onset of flow instability and the onset of nucleate boiling are calculated using the computer code PLTEMP3.MOD1 which has been developed for this analysis. The effect of fuel material (UAl x -Al, U 3 O 8 -Al and U 3 Si 2 -Al) on the peak fuel temperatures is also studied. A particular interest in the mixed core which may be constructed on the way to the use of LEU fuels, the change in the bypass flow rate due to the change in the gap between different fuel elements is investigated. (author)

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

International Nuclear Information System (INIS)

Fanjas, Y.

1981-01-01

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

Nuclear fuel storage

International Nuclear Information System (INIS)

Bevilacqua, F.

1979-01-01

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

Enriched uranium cycles in pressurized heavy water reactors

International Nuclear Information System (INIS)

Mazzola, A.

1994-01-01

A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-28

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

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)

Aldemir, T.; Turgut, H.M.; Bretscher, M.M.; Snelgrove, L.J.

1983-01-01

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 3 O 8 -Al fuel meat with a uranium density in the range 2.3 to 3.0 g/cm 3 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 U 2 O 8 fuels with densities in the 2.3 to 3.0 gU/cm 3 range. (author)

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

Uranium enrichment. Technology, economics, capacity

International Nuclear Information System (INIS)

Voigt, W.R. Jr.; Saire, D.E.; Gestson, D.K.; Peske, S.E.; Vanstrum, P.R.

1983-01-01

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+D efforts on various processes. (author)

Uranium enrichment: technology, economics, capacity

Energy Technology Data Exchange (ETDEWEB)

Voigt, Jr., W. R.; Vanstrum, P. R.; Saire, D. E.; Gestson, D. K.; Peske, S. E.

1982-08-01

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.

Uranium enrichment: technology, economics, capacity

International Nuclear Information System (INIS)

Voigt, W.R. Jr.; Vanstrum, P.R.; Saire, D.E.; Gestson, D.K.; Peske, S.E.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Sokolov, Y.

2005-01-01

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

The plutonium fuel cycles

International Nuclear Information System (INIS)

Pigford, T.H.; Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1979-07-01

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

Gadolinia experience and design for PWR fuel cycles

International Nuclear Information System (INIS)

Stephenson, L. C.

2000-01-01

The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

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

International Nuclear Information System (INIS)

Abbate, P.; Madariaga, M.R.

1990-01-01

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) [es

LEU fuel fabrication in Argentina

International Nuclear Information System (INIS)

Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.

1985-01-01

As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Modular enrichment measurement system for in-situ enrichment assay

International Nuclear Information System (INIS)

Stewart, J.P.

1976-01-01

A modular enrichment measurement system has been designed and is in operation within General Electric's Nuclear Fuel Fabrication Facility for the in-situ enrichment assay of uranium-bearing materials in process containers. This enrichment assay system, which is based on the ''enrichment meter'' concept, is an integral part of the site's enrichment control program and is used in the in-situ assay of the enrichment of uranium dioxide (UO 2 ) powder in process containers (five gallon pails). The assay system utilizes a commercially available modular counting system and a collimnator designed for compatability with process container transport lines and ease of operator access. The system has been upgraded to include a microprocessor-based controller to perform system operation functions and to provide data acquisition and processing functions. Standards have been fabricated and qualified for the enrichment assay of several types of uranium-bearing materials, including UO 2 powders. The assay system has performed in excess of 20,000 enrichment verification measurements annually and has significantly contributed to the facility's enrichment control program

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

International Nuclear Information System (INIS)

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

1994-01-01

The first low-enriched U 3 Si 2 -Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235 U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U 3 Si 2 -Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

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

Science.gov (United States)

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

2016-01-01

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

Economic study of fuel scenarios for a reload

International Nuclear Information System (INIS)

Ortiz S, J. J.; Castillo M, J. A.; Montes T, J. L.; Perusquia del C, R.

2014-10-01

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)

Disposal criticality analysis for aluminum-based DOE fuels

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1997-11-01

This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

76 FR 387 - Atomic Safety and Licensing Board; AREVA Enrichment Services, LLC (Eagle Rock Enrichment Facility)

Science.gov (United States)

2011-01-04

... and Licensing Board; AREVA Enrichment Services, LLC (Eagle Rock Enrichment Facility) December 17, 2010... construction and operation of a gas centrifuge uranium enrichment facility--denoted as the Eagle Rock... site at http://www.nrc.gov/materials/fuel-cycle-fac/arevanc.html . These and other documents relating...

A model to predict failure of irradiated U–Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Senor, David J.; Casella, Andrew M.

2016-12-15

Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

Fuel element gamma scanning at the Oak Ridge Research Reactor

International Nuclear Information System (INIS)

Hobbs, R.W.

1987-01-01

In January 1986, a demonstration program was begun at the Oak Ridge Research Reactor (ORR) to convert operations from high-enrichment uranium fuel to the newly developed U 3 Si 2 low-enrichment fuel. A primary program objective is to validate neutronics calculations conducted by the Reduced Enrichment in Research and Test Reactors Program at Argonne National Laboratory. Accordingly, a new method for determining core-power distribution has been developed. The method is based on gamma-ray spectroscopy measurements to determine the relative levels of 140 La in the fuel elements after each operating cycle. The measurement and data analyses are described and a comparison of measured and diffusion theory calculated values of the core-power distribution is presented in this paper

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.

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.

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

International Nuclear Information System (INIS)

Montierth, Leland M.

2016-01-01

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

High power test of low enriched UZrH

Energy Technology Data Exchange (ETDEWEB)

West, Gordon [General Atomic Co., San Diego, CA (United States)

1980-07-01

TRIGA-LEU fuel is currently undergoing high power tests in the 30 MW Oak Ridge Reactor. These tests are being funded by the Department of Energy through the RERTR program [Reduced Enrichment Research and Test Reactor program administered by Argonne National Laboratory] and began in mid-December, 1979 on a 16-rod shrouded cluster. The fuel rods are 0.51 in. 0D, clad with 0.16 in. Incoloy and the fuel length is 22 in. It is planned to test the UZrH fuel with 45, 30 and 20 wt-% U (nominal 20% enriched), to burnup values of about 50% of the contained U-235 in the 45 wt-% rods and about 40% and 35% burnup in the 30 wt-%, and 20 wt-% U fuel. It will take about 2 years of irradiation to produce the desired burnup in the 45 wt-% U fuel. Currently being tested are six 45 wt-% U and five 30 wt-% U rods. The remaining 5 rods are stainless steel dummies which were necessary to meet an operational requirement of the ORR which limits the power generation in a fuel rod to a value which would not raise the coolant temperature above the saturation level. Maximum calculated fuel rod powers were 40 kW, which would produce a fuel temperature of about 650 deg. C. The measured temperatures are about 400 deg. C and 350 deg. C for the 45 and 30 wt-% U fuel, respectively. Flow and {delta}T measurements show the cluster power generation to be about 250 kW, or about 65% of the design value. Reasons for the lower than expected power are still being evaluated and a proposal has been submitted for rearrangement of the fuel rods within the cluster to raise the powers and temperatures in the TRIGA-LEU fuel rods. (author)

A sensitivity study on neutronic properties of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

A sensitivity study on neutronic properties of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

Alternative Fuels Data Center: Wisconsin Reduces Emissions With Natural Gas

Science.gov (United States)

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Report of the Subcommittee on Domestic Uranium Enrichment

International Nuclear Information System (INIS)

1981-01-01

A report by the Subcommittee on Domestic Uranium Enrichment to the Atomic Energy Commission is described; which covers the procedure of the domestic uranium enrichment by centrifugal process up to the commercial production, reviewing the current situation in this field. Domestic uranium enrichment is important in the aspects of securing stable enrichment service, establishing sound fuel cycle, and others. As the future target, the production around the year 2000 is set at 3,000 tons SWU per year at least. The business of uranium enrichment, which is now developed in the Power Reactor and Nuclear Fuel Development Corporation, is to be carried out by private enterprise. The contents are as follows: demand and supply balance of uranium enrichment service, significance of domestic uranium enrichment, evaluation of centrifugal uranium enrichment technology, the target of domestic uranium enrichment, the policy of domestic uranium enrichment promotion. (J.P.N.)

Nuclear fuel utilization in Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, Z; Kharalampieva, Ts; Pejchinov, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1994-12-31

An assessment of fuel utilization in Kozloduy NPP units 1-6 is made on the basis of operational data obtained for a total of 62 fuel cycles. Basic characteristics of core loading and operation conditions are given. SPPS-1 and BIPR-7 codes are used to calculate assembly-wise power distributions for different full power days of a given cycle and unit. The data are compared with the measured values of these quantities. The analysis performed shows that the core loading option chosen has led to efficient fuel utilization without violation of the nuclear safety criteria. For WWER-440 (Units 1 - 4) this is expressed in effective reduction of the reactor vessel irradiation, maintaining the design duration of the fuel cycles at a reduced number of assemblies by a factor 5 - 5-10%, utilizing fuel with higher enrichment and implementing the 4-year fuel cycle. For WWER-1000 the improvements lead to: adoption of the 3-year fuel cycle utilizing fuel with 4.4% initial enrichment, implementation of improved fuel with a new type of absorbers and more effective low-leakage core loading patterns. 10 tabs., 6 figs., 7 refs.

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.

Supply of enriched uranium for research reactors

International Nuclear Information System (INIS)

Mueller, H.

1997-01-01

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

Nuclear fuel storage

International Nuclear Information System (INIS)

Bevilacqua, F.

1981-01-01

A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Neutron physics computation of CERCA fuel elements for Maria Reactor

International Nuclear Information System (INIS)

Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

2008-01-01

Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

Beta activity of enriched uranium

International Nuclear Information System (INIS)

Nambiar, P.P.V.J.; Ramachandran, V.

1975-01-01

Use of enriched uranium as reactor fuel necessitates its handling in various forms. For purposes of planning and organising radiation protection measures in enriched uranium handling facilities, it is necessary to have a basic knowledge of the radiation status of enriched uranium systems. The theoretical variations in beta activity and energy with U 235 enrichment are presented. Depletion is considered separately. Beta activity build up is also studied for two specific enrichments, in respect of which experimental values for specific alpha activity are available. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-09-15

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

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

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2008-01-01

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

The enrichment secondary market

International Nuclear Information System (INIS)

Einbund, D.R.

1986-01-01

This paper will addresses two topics: the background to the present status of the enrichment secondary market and the future outlook of the secondary market in enrichment services, and the viability of the nuclear fuel brokerage industry. These two topics are inevitably connected, as most secondary market activity, not only in enrichment but also in natural uranium, has traditionally been conducted with the participation of brokers. Therefore, the author interrelates these topics

Nuclear fuel management in JMTR

International Nuclear Information System (INIS)

Naka, Michihiro; Miyazawa, Masataka; Sato, Hiroshi; Nakayama, Fusao; Ito, Haruhiko

1999-01-01

The Japan Materials Testing Reactor (JMTR) is the largest scale materials (author)ted the fission gas release compared with the steady state opkW/l in Japan. JMTR as a multi-purpose reactor has been contributing to research and development on nuclear field with a wide variety of irradiation for performing engineering tests and safety research on fuel and component for light water reactor as well as fast breeder reactor, high temperature gas-cooled reactor etc., for research and development on blanket material for fusion reactor, for fundamental research, and for radio-isotope (RI) production. The driver nuclear fuel used in JMTR is aluminum based MTR type fuel. According to the Reduced Enrichment for Research and Test Reactors (RERTR) Program, the JMTR fuel elements had been converted from 93% high enriched uranium (HEU) fuel to 45% medium enriched uranium (MEU) fuel in 1986, and then to 20% low enriched uranium (LEU) fuel in 1994. The cumulative operation cycles until March 1999 reached to 127 cycles since the first criticality in 1968. JMTR has used 1,628 HEU, 688 MEU and 308 LEU fuel elements for these operation cycles. After these spent fuel elements were cooled in the JMTR water canal more than one year after discharged from the JMTR core, they had been transported to reprocessing plants in Europe, and then to plants in USA in order to extract the uranium remaining in the spent fuel. The JMTR spent fuel transportation for reprocessing had been continued until the end of 1988. However, USA had ceased spent fuel reprocessing in 1989, while USDOE committed to prepare an environmental review of the impacts of accepting spent fuels from foreign research reactors. After that, USDOE decided to implement a new acceptance policy in 1996, the spent fuel transportation from JMTR to Savannah River Site was commenced in 1997. It was the first transportation not only in Japan but in Asia also. Until resuming the transportation, the spent fuel elements stored in JMTR

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.

The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

International Nuclear Information System (INIS)

Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

2006-01-01

Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

MOX fuel assembly and reactor core

International Nuclear Information System (INIS)

Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

1998-01-01

The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors' spent fuel

International Nuclear Information System (INIS)

1994-01-01

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE's Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE's efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE's activities in taking back spent fuel

Application research of improved 235U enrichment meter

International Nuclear Information System (INIS)

Liu Daming; Wu Xin; Lu Zhao; Tang Peijia; Lu Feng; Wang Yunmei

1998-01-01

A prototype 235 U enrichment meter based on NaI(Tl) γ spectroscopy is improved and it works under the principle of that the enrichment of 235 U is proportional to the radioactivity of 185 keV γ-ray when the sample is thick infinitely. The data of radioactivity from 235 U can be collected by a notebook computer and the interface control software is written using C++ language. The meter was tested and calibrated using standard fuel rods in fuel fabrication plant. For single fuel rod, the measured value of 235 U enrichment is agreeable with declared value within-1.0%-2.8%

Swelling Estimation of Multi-wire U-Mo Monolithic Fuel for HANARO Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon-Sang; Ryu, Ho-Jin; Park, Jong-Man; Oh, Jong-Myeong; Kim, Chang-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm - 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix. In this study temperature calculations and a swelling estimation of a multi-wire monolithic fuel were carried out. Also the results of a post irradiation analysis of this fuel will be introduced.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

High enrichment to low enrichment core's conversion. Accidents analysis

International Nuclear Information System (INIS)

Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

1990-01-01

This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. 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 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-21

Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd{sub 2}O{sub 3}) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

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

International Nuclear Information System (INIS)

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

2013-01-01

Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd 2 O 3 ) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241 AmLi (α,n) interrogation source strength of 5.7×10 4 s −1 . Furthermore, the calibration range of the new collar has been extended to verify 235 U content in variable PWR fuel designs in the presence of up to

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

Energy Technology Data Exchange (ETDEWEB)

Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

2016-06-01

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

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

Energy Technology Data Exchange (ETDEWEB)

Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

Low enrichment of uranium in the light of the nuclear weapon problem

International Nuclear Information System (INIS)

Barstad, G.

1979-09-01

A difficult problem in the immediate future will be to direct civil nuclear technology in such a way that the ability to produce nuclear weapons by additional countries is prevented. There are two main problems. First, enrichment plants can be used to produce high enriched uranium, which can be used in nuclear weapons, as well as low enriched reactor fuel. Second, plutonium produced during reactor operation can be used as nuclear weapon material, as well as for nuclear fuel. The problem discussed here is particularly the development of an enrichment process which is economic for low enriched reactor fuel, but which may not easily be adapted to produce high enriched uranium. (JIW)

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

Science.gov (United States)

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

2014-08-01

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

Use of enriched uranium in Canada's power reactors

International Nuclear Information System (INIS)

Dormuth, K.W.; Jackson, D.P.

2011-01-01

Recent trends in Canadian nuclear power reactor design and proposed development of nuclear power in Canada have indicated the possibility that Canada will break with its tradition of natural uranium fuelled systems, designed for superior neutron economy and, hence, superior uranium utilization. For instance, the Darlington B new reactor project procurement process included three reactor designs, all employing enriched fuel, although a natural uranium reactor design was included at a late stage in the ensuing environmental assessment for the project as an alternative technology. An evaluation of the alternative designs should include an assessment of the environmental implications through the entire fuel cycle, which unfortunately is not required by the environmental assessment process. Examples of comparative environmental implications of the reactor designs throughout the fuel cycle indicate the importance of these considerations when making a design selection. As Canada does not have enrichment capability, a move toward the use of enriched fuel would mean that Canada would be exporting natural uranium and buying back enriched uranium with value added. From a waste management perspective, Canada would need to deal with mill, refinery, and conversion tailings, as well as with the used fuel from its own reactors, while the enrichment supplier would retain depleted uranium with some commercial value. On the basis of reasoned estimates based on publicly available information, it is expected that enrichment in Canada is likely to be more profitable than exporting natural uranium and buying back enriched uranium. Further, on the basis of environmental assessments for enrichment facilities in other countries, it is expected that an environmental assessment of a properly sited enrichment facility would result in approval. (author)

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

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Weeks, J.

1994-01-01

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

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

Energy Technology Data Exchange (ETDEWEB)

Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

2014-12-15

Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

Advanced fuel cycles for WWER-1000 reactors

International Nuclear Information System (INIS)

Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

2007-01-01

Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

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

Directory of Open Access Journals (Sweden)

Varese Giovanna C

2010-02-01

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

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

Science.gov (United States)

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

2010-02-16

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

French LEU fuel for research reactor with emphasis on the Osiris experience of core conversion and reactor operation with the new fuel

International Nuclear Information System (INIS)

Cerles, J.-M.

1981-09-01

One of the various activities carried out in France concerned with the design, fabrication and development of nuclear fuels was the development by the CEA of a plate type fuel (Caramel fuel). A Caramel fuel element is in the form of a plate consisting of two tight covering zircaloy sheets in which the UO 2 platelets are confined themselves within the network of a zircaloy grid. The plane geometry provides an effective means of overcoming the drawback of poor uranium oxide conductivity, and makes it possible to combine high specific power with low fuel temperature. The chief advantages of this fuel are the following: it is a very low enriched fuel. It can be used in research reactors demanding high volumetric powers and neutron fluxes, with a required enrichment significantly lower than 20% 235 U. The difference between the densities of UO 2 matrix and U-Al, 10.3 and 1.6 g/cm respectively, leads to a higher uranium charge, making it possible to reduce the enrichment to between 3 and 10%. Owing to fuel dispersion, any loss of tightness only puts a small amount of fissile material in contact with the coolant, thus limiting any contamination of the primary circuit. Another safety factor is the operating temperature, which is considerably lower than the temperature at which fission gases are liberated

Nuclear fuel elements and assemblies

International Nuclear Information System (INIS)

Saito, Shozo; Maki, Hideo.

1982-01-01

Purpose: To facilitate the attainment of the uranium enrichment or gadolinia enrichment of a pellet filled in a fuel element. Constitution: The axial length of a pellet filled in a fuel element is set to predetermined sizes according to the uranium enrichment factor, gadolinia enrichment or their combination. Thus, the uranium enrichment factor or gadolinia enrichment can be identified by attaining the axial length of the pellet by using such a pellt. (Kamimura, M.)

Update on international uranium and enrichment supply

International Nuclear Information System (INIS)

Cleveland, J.M.

1987-01-01

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

Fuel and fuel cycles with high burnup for WWER reactors

International Nuclear Information System (INIS)

Chernushev, V.; Sokolov, F.

2002-01-01

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)

Analysis of some fuel characteristics deviations and their influence over WWER-440 fuel cycle design

International Nuclear Information System (INIS)

Stoyanova, I.; Kamenov, K.

2001-01-01

The aim of this study is to estimate the influence of some deviations in WWER-440 fuel assemblies (FA) characteristics upon fuel core design. A large number of different fresh fuel assemblies with enrichment of 3.5 t % are examined related to the enrichment, mass of initial metal Uranium and assembly shroud thickness. Infinite multiplication factor (Kinf) in fuel assembly has been calculated by HELIOS spectral code for basic assembly and for different FA with deviation of a single parameter. The effects from single parameter deviation (enrichment) and from two parameter deviations (enrichment and wall thickness) on the neutron-physics characteristics of the core are estimated for different fuel assemblies. Relatively week burnup dependence on Kinf is observed as result of deviation in the enrichment of the fuel and in the wall thickness of the assembly. An assessment of a FA single and two parameter deviations effects on design fuel cycle duration and relative power peaking factor is also considers in the paper. As a final conclusion can be settled that the maximum relative shortness of fuel cycle can be observed in the case of two FA parameters deviations