WorldWideScience

Sample records for foreign-owned enriched uranium

  1. Transportation of foreign-owned enriched uranium from the Republic of Georgia. Environmental assessment for Project Partnership

    International Nuclear Information System (INIS)

    1998-01-01

    The Department of Energy (DOE) Office of Nonproliferation and National Security (NN) has prepared a classified environmental assessment to evaluate the potential environmental impact for the transportation of 5.26 kilograms of enriched uranium-235 in the form of nuclear fuel, from the Republic of Georgia to the United Kingdom. The nuclear fuel consists of primarily fresh fuel, but also consists of a small quantity (less than 1 kilogram) of partially-spent fuel. Transportation of the enriched uranium fuel would occur via US Air Force military aircraft under the control of the Defense Department European Command (EUCOM). Actions taken in a sovereign nation (such as the Republic of Georgia and the United Kingdom) are not subject to analysis in the environmental assessment. However, because the action would involve the global commons of the Black Sea and the North Sea, the potential impact to the global commons has been analyzed. Because of the similarities in the two actions, the Project Sapphire Environmental Assessment was used as a basis for assessing the potential impacts of Project Partnership. However, because Project Partnership involves a small quantity of partially-spent fuel, additional analysis was conducted to assess the potential environmental impacts and to consider reasonable alternatives as required by NEPA. The Project Partnership Environmental Assessment found the potential environmental impacts to be well below those from Project Sapphire

  2. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

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

  3. 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

  4. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Thomas, D.C.; Gagne, R.W.

    1978-01-01

    The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

  5. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Gagne, R.W.; Thomas, D.C.

    1977-01-01

    The status of existing uranium enrichment contracts in the US is reviewed and expected natural uranium requirements for existing domestic uranium enrichment contracts are evaluated. Uncertainty in natural uranium requirements associated with requirements-type and fixed-commitment type contracts is discussed along with implementation of variable tails assay

  6. 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

  7. Uranium enrichment: an overview

    Energy Technology Data Exchange (ETDEWEB)

    Cazalet, J.

    1995-12-31

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

  8. Uranium enrichment: an overview

    International Nuclear Information System (INIS)

    Cazalet, J.

    1995-01-01

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

  9. Uranium Conversion & Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-06

    The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U3O8 yellowcake into UF6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

  10. 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.)

  11. South Australia, uranium enrichment

    International Nuclear Information System (INIS)

    1976-02-01

    The Report sets out the salient data relating to the establishment of a uranium processing centre at Redcliff in South Australia. It is conceived as a major development project for the Commonwealth, the South Australian Government and Australian Industry comprising the refining and enrichment of uranium produced from Australian mines. Using the data currently available in respect of markets, demand, technology and possible financial return from overseas sales, the project could be initiated immediately with hexafluoride production, followed rapidly in stages by enrichment production using the centrifuge process. A conceptual development plan is presented, involving a growth pattern that would be closely synchronised with the mining and production of yellowcake. The proposed development is presented in the form of an eight-and-half-year programme. Costs in this Report are based on 1975 values, unless otherwise stated. (Author)

  12. Advanced uranium enrichment processes

    International Nuclear Information System (INIS)

    Clerc, M.; Plurien, P.

    1986-01-01

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

  13. Uranium enriched granites in Sweden

    International Nuclear Information System (INIS)

    Wilson, M.R.; Aakerblom, G.

    1980-01-01

    Granites with uranium contents higher than normal occur in a variety of geological settings in the Swedish Precambrian, and represent a variety of granite types and ages. They may have been generated by (1) the anatexis of continental crust (2) processes occurring at a much greater depth. They commonly show enrichement in F, Sn, W and/or Mo. Only in one case is an important uranium mineralization thought to be directly related to a uranium-enriched granite, while the majority of epigenetic uranium mineralizations with economic potential are related to hydrothermal processes in areas where the bedrock is regionally uranium-enhanced. (Authors)

  14. AEC determines uranium enrichment policy

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The Advisory Committee on Uranium Enrichment of the Atomic Energy Commission (AEC) has submitted a report to AEC chairman concerning the promotion of the introduction of advanced material, high performance centrifuges to replace conventional metallic drum centrifuges, and the development of next generation advanced centrifuges. The report also called for the postponement until around 1997 of the decision whether the development should be continued or not on atomic vapor laser isotope separation (AVLIS) and molecular laser isotope separation (MLIS) processes, as well as the virtual freezing of the construction of a chemical process demonstration plant. The report was approved by the AEC chairman in August. The uranium enrichment service market in the world will continue to be characterized by oversupply. The domestic situation of uranium enrichment supply-demand trend, progress of the expansion of Rokkasho enrichment plant, the trend in the development of gas centrifuge process and the basic philosophy of commercializing domestic uranium enrichment are reported. (K.I.)

  15. Uranium enrichment by gas centrifuge

    International Nuclear Information System (INIS)

    Heriot, I.D.

    1988-01-01

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

  16. The isotopic enrichment of uranium in 1979

    International Nuclear Information System (INIS)

    Baron, M.

    1979-01-01

    The Eurodif uranium enrichment plant built on the Tricastin site is described. The uranium isotope separation plants in service abroad are presented. The main characteristics of the international enrichment market are defined [fr

  17. Environmental Development Plan: uranium enrichment

    International Nuclear Information System (INIS)

    1979-09-01

    This Environmental Development Plan identifies and examines the environmental, health, safety, and socioeconomic concerns and corresponding requirements associated with the DOE research, development, demonstration, and operation of the Uranium Enrichment program, including the gaseous diffusion process, the centrifuge process, centrifuge rotor fabrication, and related research and development activities

  18. Uranium enrichment plans and policies

    International Nuclear Information System (INIS)

    Schwennesen, J.L.

    1981-01-01

    Significant progress has been made in US efforts to expand its enrichment capacity. The Cascade Improvement Program (CIP) and Cascade Upgrading Program (CUP) are now complete at Oak Ridge and Paducah and almost complete at Portsmouth. Considerable progress has also been made in constructing the Gas Centrifuge Enrichment Plant (GCEP), and physical construction of the first process building is well under way. Current plans are to have two process buildings on-line by 1989 with the remaining six buildings to be added sequentially as needed to meet demand. The status of DOE enrichment services contracts is essentially unchanged from that reported at last year's seminar. The OUEA latest forecast of nuclear power growth, however, is considerably lower than reported last year, although a leveling trend is becoming apparent. The Variable Tails Assay Option (VTAO) of the AFC contract was made available for the third time for FY 1983. The DOE inventories of natural uranium still remain high. The Department of Energy will dispose of this material by using it for Government programs and for enrichment plant operations. It appears that Government inventories of uranium are adequate through at least the mid-1990s. It remains DOE policy not to dispose of its natural uranium stocks through direct sales in the marketplace, except for very small quantities or if an emergency situation would exist and all reasonable attempts had been made, without success, to obtain natural uranium from commercial sources. Finally, with regard to DOE plans on future transaction tails assays, it still appears likely that the current 0.20 percent uranium-235 reference tails assay will be maintained until well into the 1990s, at which time it might be increased up to 0.25 percent uranium-235

  19. 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

  20. 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

  1. Uranium enrichment. 1980 annual report

    International Nuclear Information System (INIS)

    1981-05-01

    This report contains data and related information on the production of enriched uranium at the gaseous diffusion plants and an update on the construction and project control center for the gas centrifuge plant. Power usage at the gaseous diffusion plants is illustrated. The report contains several glossy color pictures of the plants and processes described. In addition to gaseous diffusion and the centrifuge process, three advanced isotope separation process are now being developed. The business operation of the enrichment plants is described; charts on revenue, balance sheets, and income statements are included

  2. Proliferation implications of uranium enrichment technologies

    International Nuclear Information System (INIS)

    1979-02-01

    A summary chart is developed comparing the difficulties inherent in the production of highly enriched uranium via various enrichment technologies, on the assumption that the necessary technology is not available and must be acquired. Production of highly enriched uranium by batch recycling in existing facilities is also discussed, and some safeguards measures are suggested

  3. Uranium enrichment in the United States

    International Nuclear Information System (INIS)

    Hill, J.H.; Parks, J.W.

    1975-01-01

    History, improvement programs, status of electrical power availability, demands for uranium enrichment, operating plan for the U. S. enriching facilities, working inventory of enriched uranium, possible factors affecting deviations in the operating plan, status of gaseous diffusion technology, status of U. S. gas centrifuge advances, transfer of enrichment technology, gaseous diffusion--gas centrifuge comparison, new enrichment capacity, U. S. separative work pricing, and investment in nuclear energy are discussed. (LK)

  4. Blending UNH streams of different uranium enrichments

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, B.F.

    1958-04-18

    The cost and feasibility of blending UNH streams of different uranium enrichments at HAPO was evaluated on a preliminary basis. Cases studied were blending 37.5% enriched UNH with depleted E metal UNH to yield a 0.947% enriched end product, and blending depleted E metal UNH with the depleted natural uranium to yield 0.7115% enriched end product. A reasonable degree of feasibility is indicated for such a blend program at HAPO.

  5. Perspectives for the uranium enrichment in Brazil

    International Nuclear Information System (INIS)

    Senna, J.G.S.M.

    1991-01-01

    Through an analysis of the electrical energy future in Brazil, the needs for enriched uranium are discussed, and therefore the importance of developing local capability for self-production. A description of the production processes that are well established is given first, then the analysis itself is performed and finally a visualization of the International Market for enriched uranium is shown. (author)

  6. EURODIF: the uranium enrichment by gaseous diffusion

    International Nuclear Information System (INIS)

    Rougeau, J.P.

    1981-01-01

    During the seventies the nuclear power programme had an extremely rapid growth rate which entailed to increase the world uranium enrichment capacity. EURODIF is the largest undertaking in this field. This multinational joint venture built and now operates and enrichment plant using the gaseous diffusion process at Tricastin (France). This plant is delivering low enriched uranium since two years and has contracted about 110 million SWU's till 1990. Description, current activity and prospects are given in the paper. (Author) [pt

  7. Current perspective of the uranium enrichment market

    International Nuclear Information System (INIS)

    Laughon, K.O.

    1986-01-01

    Over the past several years, developments in the uranium enrichment market have required the Department of Energy (DOE) to make a number of changes in the U.S. enrichment enterprise. These changes have been made to allow DOE to conduct our enrichment business so as to be more responsive to changing market forces. Needless to say, some of these changes have been difficult, but they have been necessary if they are to conduct a healthy and competitive uranium enrichment business in the United States. This paper discusses several topics, including: The Uranium Enrichment Market, Utility Services (US) Contracts, Reduced Prices, Incentive Pricing, Better Customer Services, and Advanced Technology. In addition to these topics, information is provided on the recent court action regarding the US Contracts and the viability finding on the uranium mining industry

  8. Uranium enrichment by diffusion process

    International Nuclear Information System (INIS)

    Petit, J.F.

    1977-01-01

    A brief history of the process is given, and existing facilities are described, and the possibility of this process meeting the needs for enriched uranium in the future is discussed. It is concluded that, since other separation techniques have not yet been proven to be competitive, the gaseous diffusion process must be relied on to meet most of these needs. However, because of the high energy consumption required by this process, attempts to improve the technology must be continued. Is the gaseous diffusion process able to meet this challenge. Considering the technical viewpoint, there is no doubt about it. From the economical standpoint, this process is mainly characterized by an important energy consumption and the necessity to build large plants, thus leading to an important investment, at least for the first plant. Other processes have been developed with a view to reduce both the energy and capital needs. However, in spite of continuous studies and technological progress no process has yet been proven to be competitive. Large increments in capacities are still expected to come from the gaseous diffusion and several projects taking into acount the improvements in the flexibility, automatization, reliability and a reduced investment, will be analysed. Coupling of new facilities to existing plants has already proven to be of great interest. This situation explains why further investigations concerning the gaseous diffusion are being carried on, together with the study of new processes [fr

  9. Enriched uranium sales: effect on supply industry

    International Nuclear Information System (INIS)

    Andersen, R.K.

    1985-01-01

    The subject is covered in sections: introduction (combined effect of low-enriched uranium (LEU) inventory sales and utility services enrichment contract terms); enrichment market overview; enrichment market dynamics; the reaction of the US Department of Energy; elimination of artificial demand; draw down of inventories; purchase and sale of LEU inventories; tails assay option; unfulfilled requirements for U 3 O 8 ; conclusions. (U.K.)

  10. Mo-99 from low-enriched uranium

    International Nuclear Information System (INIS)

    Cols, Hector J.; Cristini, Pablo R.; Manzini, Alberto C.

    2000-01-01

    The use of low-enriched uranium aluminide targets for the production of fission Mo-99 is presented. Former tests on uranium silicide targets are also reviewed. Tests of dissolution of aluminide miniplates in alkaline medium were performed, showing interesting results for the replacement of current HEU targets. (author)

  11. Uranium enrichment services in the United States

    International Nuclear Information System (INIS)

    Jelinek, P.; Lenders, M.

    1994-01-01

    The United States of America is the world's largest market for uranium enrichment services. After the disintegration of the Soviet Union, Russian uranium is entering the world market on an increasing scale. The U.S. tries to protect its market and, in this connection, also the European market from excessive price drops by taking anti-dumping measures. In order to become more competitive, American companies have adapted modern enrichment techniques from Europe. European - U.S. joint ventures are to help, also technically and economically, to integrate military uranium, accumulating as a consequence of worldwide disarmament, into the commercial fuel cycle for the peaceful use of nuclear power. (orig.) [de

  12. Uranium enrichment: a vital new industry

    International Nuclear Information System (INIS)

    1975-10-01

    The energy problem facing the nation and the need for nuclear power are pointed out. Uranium enrichment and the demand for it are discussed. Reasons for, and obstacles to, private enrichment are outlined. The President's plan (including the Nuclear Fuel Assurance Act) is summarized

  13. The evolution of the enriched uranium markets

    International Nuclear Information System (INIS)

    Arnaiz, J.; Moleres, C.; Tarin, F.

    2004-01-01

    This paper deals with the evolution of the enriched uranium component markets (uranium concentrates, conversion and enrichment), starting with the situation of historically low prices that occurred during 2000. The situation that has been reached as on December 2003, when the concentrates and conversion markets were 44% and 70% (current US$) respectively, and the enrichment prices 30%, higher, is analysed. Finally, the negative impact of the 90's depressed prices, due to abundant alternative sources of uranium components, on the primary production of all three components and, as a conclusion, the impact of the new situation on the transport logistics, and the need of appropriate economic conditions to make the future primary production sustainable, is commented. (Author)

  14. Distribution of uranium supply and enrichment

    International Nuclear Information System (INIS)

    Bamford, F.W.

    1982-01-01

    Uranium supply and demand is examined from the perspective of companies in the uranium hexafluoride (UF6) conversion business whose main interest is their sources of uranium supply and UF6 destinations because of transportation costs. Because of the variations in yellowcake transport, charges for conversion, and UF6 transport costs, most converters don't have standard prices. Companies try to look ahead to determine patterns of supplies and delivery points when they analyze the market and estimate future prices. Market analyses must take into account the purchasing policies of utilities around the world. The presentation shows North America supplying about 40% of world uranium, with about 13% of the enrichment done elsewhere. It also shows North American converters getting 53% of the business, but that will require importing uranium from outside North America. 6 tables

  15. Uranium enrichment: heading for the abyss

    International Nuclear Information System (INIS)

    Norman, C.

    1983-01-01

    This article discusses the federal government's $2.3 billion a year business enriching uranium for nuclear power plants which is heading toward a major crisis. Due to miscalculations by the Department of Energy, it is caught with billions of dollars of construction in progress just as projected demand for enriched uranium is decreasing. At the center of the controversy is the Gas Centrifuge Plant at Portsmouth, Ohio - estimated to cost $10 billion dollars. A review of how DOE got into this situation and how they plan to solve it is presented

  16. Profile of World Uranium Enrichment Programs - 2007

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    2007-11-01

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

  17. The supply of the European community countries with enriched uranium

    International Nuclear Information System (INIS)

    1975-02-01

    A discussion is given of a survey regarding the supply of enriched uranium to the countries of the European Community. Costs of enriched uranium imports were not available but import values were calculated using world market prices. (R.L.)

  18. Feasibility of uranium enrichment in Australia

    International Nuclear Information System (INIS)

    1979-10-01

    The Council considered that provided the balance between costs and markets was found to be acceptable, there was no valid reason against the Government proceeding with a study on the feasibility of, and perhaps participating in the establishment of a commercial uranium enrichment industry in Australia. Areas covered include technical expertise and industrial structure in Australia, environmental aspects and safeguards

  19. Topical papers on uranium conversion and enrichment

    International Nuclear Information System (INIS)

    Uranium conversion and enrichment are discussed in 5 papers by representatives of the USA, Great Britain and Switzerland. The state of the art is reviewed, and future prospects are given. Supply assurance is directly related to the necessary production capacities and the supply agreements

  20. Enriched uranium recovery at Los Alamos

    International Nuclear Information System (INIS)

    Herrick, C.C.

    1984-01-01

    Graphite casting scrap, fuel elements and nongraphite combustibles are calcined to impure oxides. These materials along with zircaloy fuel elements and refractory solids are leach-dissolved separately in HF-HNO 3 acid to solubilize the contained enriched uranium. The resulting slurry is filtered and the clear filtrate (to which mineral acid solutions bearing enriched uranium may be added) are passed through solvent extraction. The solvent extraction product is filtered, precipitated with H 2 O 2 and the precipitate calcined to U 3 O 8 . Metal is made from U 3 O 8 by conversion to UO 2 , hydrofluorination and reduction to metal. Throughput is 150 to 900 kg uranium per year depending on the type of scrap

  1. Long term assurance of supply of uranium enrichment

    International Nuclear Information System (INIS)

    1978-01-01

    After elaborating a number of key questions on uranium enrichment, the representatives of 10 countries and of the EC commission present their answers. Attention is paid to the assurance of uranium supply, to uranium enrichment, market trends and flexibility in enrichment agreements

  2. How is uranium supply affecting enrichment?

    International Nuclear Information System (INIS)

    Steve Kidd

    2007-01-01

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

  3. Uranium enrichment management review: summary of analysis

    International Nuclear Information System (INIS)

    1981-01-01

    In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices

  4. Multinational uranium enrichment in the Middle East

    International Nuclear Information System (INIS)

    Ahmad, Ali; Salahieh, Sidra; Snyder, Ryan

    2017-01-01

    The Joint Comprehensive Plan of Action (JCPOA) agreed to by Iran and the P5+1 in July 2015 placed restrictions on Iran’s nuclear program while other Middle Eastern countries– Egypt, Jordan, Saudi Arabia, Turkey, and the United Arab Emirates–are planning to build their own nuclear power plants to meet increasing electricity demands. Although the JCPOA restricts Iran's uranium enrichment program for 10–15 years, Iran's neighbors may choose to develop their own national enrichment programs giving them a potential nuclear weapons capability. This paper argues that converting Iran's national enrichment program to a more proliferation-resistant multinational arrangement could offer significant economic benefits–reduced capital and operational costs–due to economies of scale and the utilization of more efficient enrichment technologies. In addition, the paper examines policy aspects related to financing, governance, and how multinational enrichment could fit into the political and security context of the Middle East. A multinational enrichment facility managed by regional and international partners would provide more assurance that it remains peaceful and could help build confidence between Iran and its neighbors to cooperate in managing other regional security challenges. - Highlights: • Freezing Iran's nuclear program is an opportunity to launch joint initiatives in ME. • A joint uranium enrichment program in the Middle East offers economic benefits. • Other benefits include improved nuclear security and transparency in the region.

  5. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF 6 and a (2) blend the pure HEU UF 6 with diluent UF 6 to produce LWR grade LEU-UF 6 . The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry

  6. Criticality analysis in uranium enrichment plant

    International Nuclear Information System (INIS)

    Okamoto, Tsuyoshi; Kiyose, Ryohei

    1977-01-01

    In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

  7. Detection of uranium enrichment activities using environmental monitoring techniques

    International Nuclear Information System (INIS)

    Belew, W.L.; Carter, J.A.; Smith, D.H.; Walker, R.L.

    1993-01-01

    Uranium enrichment processes have the capability of producing weapons-grade material in the form of highly enriched uranium. Thus, detection of undeclared uranium enrichment activities is an international safeguards concern. The uranium separation technologies currently in use employ UF 6 gas as a separation medium, and trace quantities of enriched uranium are inevitably released to the environment from these facilities. The isotopic content of uranium in the vegetation, soil, and water near the plant site will be altered by these releases and can provide a signature for detecting the presence of enriched uranium activities. This paper discusses environmental sampling and analytical procedures that have been used for the detection of uranium enrichment facilities and possible safeguards applications of these techniques

  8. Detection of uranium enrichment activities using environmental monitoring techniques

    Energy Technology Data Exchange (ETDEWEB)

    Belew, W.L.; Carter, J.A.; Smith, D.H.; Walker, R.L.

    1993-03-30

    Uranium enrichment processes have the capability of producing weapons-grade material in the form of highly enriched uranium. Thus, detection of undeclared uranium enrichment activities is an international safeguards concern. The uranium separation technologies currently in use employ UF{sub 6} gas as a separation medium, and trace quantities of enriched uranium are inevitably released to the environment from these facilities. The isotopic content of uranium in the vegetation, soil, and water near the plant site will be altered by these releases and can provide a signature for detecting the presence of enriched uranium activities. This paper discusses environmental sampling and analytical procedures that have been used for the detection of uranium enrichment facilities and possible safeguards applications of these techniques.

  9. Uranium enrichment by centrifuge in Japan

    International Nuclear Information System (INIS)

    Watanabe, T.; Murase, T.

    1977-01-01

    The demand for enriched uranium is on the increase with nuclear power capacity in which the LWR predominates and is estimated to exceed the supply from the present facilities in the world in less than ten years. Therefore, the basic strategy for enriched uranium is investigated on the following three-point long-range program in Japan: 1. To continue negotiations to extend the current allocation by the long-term contract; 2. To seek active participation in international enrichment projects; and 3. To make efforts to develop uranium enrichment technology and to construct inland facilities. On this basis, a vigorous development program of gas centrigue process for industrialization was launched out in 1972 as a national project. Ever since substantial progress in this field has been made and development works have been increased year after year. At present, a concrete plan of a pilot plant is taking shape. Up to now, several types of centrifuges were developed, of which some were completed as prototype models, and subjected to life tests and also to extensive earthquake-resistivity tests for the characteristics of Japanese geological condition. An enrichment plant is composed of so many centrifuges that the installation and piping system of centrifuges is an important factor which has an effect on plant economy and reliability. Two types of the experimental cascade were constructed in Japan. One has been in operation since 1973, and the other since 1975. Valuable empirical data have been accumulated on cascade characteristics, maintenance scheme and so on. It will be important for the coming plants to have a flexibility to escalation of labor and energy cost, or to variation of the separative work requirement and further. An economic prospect of centrifuge enrichment process is presented

  10. Uranium enrichment and the environment

    International Nuclear Information System (INIS)

    Wing, J.F.

    1980-01-01

    The three gaseous diffusion plants comprise one of the largest industrial complexes in the free world. Since all of the enrichment is done inside a very large plumbing maze, the control of radioactivity does not present a serious problem. Conventional, nonexotic engineering and administrative measures adequately control the minor levels of radioactivity associated with support activities such as equipment decontamination and maintenance. The treatment and control of chemical waste streams to comply with Federal and state regulations has required the commitment of 47 million dollars since 1974. Through 1982, an additional 84 million dollars may be required. The high cost is not a reflection of initially poor conditions but is rather the result of providing large treatment systems to meet very low discharge limits. Examples that will be discussed include airborne particulate removal, recirculating cooling water treatment, and hazardous waste disposal concepts

  11. Uranium enrichment. Industrial and commercial aspect

    International Nuclear Information System (INIS)

    Lamorlette, G.

    1983-01-01

    The uranium enrichment, a key stage in the fuel cycle of light-water nuclear power stations, applies sophisticated and protected techniques in installations on a very large scale. This article shows how there was a sudden change from a monopoly position in production to a severe competition in a market which is depressed today but offers good prospects for the future. It indicates how the enrichment industrialist have adapted themselves to the fluctuations of the demand, while safeguarding the reliability of the rendered service and the necessary security of supplies for the proper development of the nuclear electric power [fr

  12. Profile of World Uranium Enrichment Programs-2009

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    2009-04-01

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

  13. Uranium enrichment export control guide: Gaseous diffusion

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document was prepared to serve as a guide for export control officials in their interpretation, understanding, and implementation of export laws that relate to the Zangger International Trigger List for gaseous diffusion uranium enrichment process components, equipment, and materials. Particular emphasis is focused on items that are especially designed or prepared since export controls are required for these by States that are party to the International Nuclear Nonproliferation Treaty.

  14. R and D on laser uranium enrichment

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    An AEC Advisory Committee on Uranium Enrichment has completed investigations into the actual condition of laser isotope separation. The working group set up for the purpose has issued a report on the series of investigations made on its development and measures for promoting it. The report says that the development of the process in Japan is at a fundamental stage. Noting that further efforts are needed before its future can be predicted, the report proposes a cource of research and development for the immediate future. For the atomic vapor laser isotope separation (AVLIS), government organizations are engaged in data base buildup and conducting basis engineering tests, and Japan Atomic Energy Research Institute will consider the re-enrichment of uranium recovered from reprocessing. Non-governmental unions of researchers will promote the combination of copper-vapor laser and dye laser. For the molecular laser isotope separation (MLIS), the Institute of Physical and Chemical Research will take up studies with the cooperation of the Power Reactor and Nuclear Fuel Development Corporation. In chapters covering the philosophy of laser uranium enrichment technology development, the report deals with its significance, actual conditions and tasks, and goals and measures for its promotion. (Nogami, K.)

  15. URENCO. Uranium enrichment with advanced technology

    International Nuclear Information System (INIS)

    2011-01-01

    URENCO Deutschland is a subsidiary of URENCO Enrichment Company Limited, an international enterprise founded in 1970 in the State Treaty of Almelo, which offers uranium enrichment for nuclear power plants all over the world with the use of advanced technology. URENCO facilities at present are operated in the United Kingdom, the Netherlands, USA, and in Germany. The German URENCO location is Gronau, Westphalia, where cascades have been in operation since 1985 using centrifuge technology to enrich nuclear fuel to up to 5% uranium-235. The URENCO Group supplies nuclear power plants in Europe and overseas countries with a world market share, at present, of more than 25% with a rising tendency. The first uranium separation plant in Gronau (UTA-1) attained its full separation performance of 1,800 t USW/a in late 2005. In February 2005, construction and operation of another plant had been licensed, which can raise the aggregate capacity on site to 4,500 t USW per annum. Construction of the new plant (UTA-2) was begun in summer 2005. UTA-2 will use the latest, most powerful URENCO centrifuge. URENCO has more than 3,500 visitors a year at its German location alone, thus demonstrating its pro-active information policy and offering to the public a maximum of opportunities to acquire information by attending presentations and tours of the plant. (orig.)

  16. 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

  17. Uranium enrichment management review. Final report

    International Nuclear Information System (INIS)

    Ellett, J.D.; Rieke, W.B.; Simpson, J.W.; Sullivan, P.E.

    1980-01-01

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

  18. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

  19. Comments on Smith Barney's uranium enrichment analysis

    International Nuclear Information System (INIS)

    Rezendes, V.S.

    1990-07-01

    In a May 1990 report, Smith Barney, Harris Upham and Co. concluded that DOE's uranium enrichment program should be restructured as a government corporation; all past costs have been recovered, and DOE's customers have been overcharged about $1.2 billion; the government should retain responsibility for environment and decommissioning costs associated with enriched uranium production before the corporation's formation; and at some future time the corporation could be sold to the private sector. This report agrees with Smith Barney's recommendation to restructure the enrichment program as a government corporation, but disagrees that DOE's customers have paid for all past costs. According to the author, Smith Barney did not identify the total environmental or decommissioning costs between the government and the corporation. Since these costs are largely undefined, but could amount to billions, Congress should immediately require the program to begin setting aside funds for these costs. DOE estimates that government purchases are responsible for 50 percent of the decommissioning costs; therefore, the government should share these costs by matching the corporation's fund contributions. This requirement should continue until the existing plants have been decommissioned

  20. Evaluation of uranium-enrichment processes

    International Nuclear Information System (INIS)

    Vanstrum, P.R.; Wilcox, W.J. Jr.; McGill, R.M.

    1982-01-01

    The purpose of this paper is to outline some of the methods used in evaluating uranium enrichment processes. The paper shows how one can choose among these many processes and what features are crucial in deciding between them. These features can be grouped into technical and economic factors. The technical factors include separation factor, throughput, inventory of material, and specific power requirement. The economic factors are: (1) capital cost of plant and supporting facilities; (2) operating costs including maintenance costs; and (3) power costs. Besides these essential factors, other factors may also need to be considered. These are: (1) potential for further process improvements; (2) reliability of the process equipment; (3) difficulty of manufacturing and handling the process medium; (4) process flexibility; (5) scaling factors; (6) social factors; (7) political factors. The paper illustrates how these basic technical and economic evaluation factors are applied to real enrichment methods. The gaseous diffusion process is used as a reference process and compared with the following processes: gas centrifuge; advanced isotope separation; thermal diffusion; fractional distillation; mass spectrograph; and mass diffusion. Another point that needs to be considered is that some evaluations of uranium enrichment processes change over a period of time because of advances in technology or changes in economic climate

  1. The case for enrichment of uranium in Australia

    International Nuclear Information System (INIS)

    George, D.W.

    1981-01-01

    Information is presented on the number of nuclear power plants in operation and under construction and on the extent of the use of uranium. The case for enrichment of uranium in Australia is then considered in detail and the status of feasbility studies being carried out is discussed. Arguments to support an enrichment industry include: the need for additional enrichment capacity; added value; potential profitability; increased employment and industrial opportunities; and retention of depleted uranium

  2. Surplus Highly Enriched Uranium Disposition Program plan

    International Nuclear Information System (INIS)

    1996-10-01

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

  3. Active interrogation of highly enriched uranium

    Science.gov (United States)

    Fairrow, Nannette Lea

    Safeguarding special nuclear material (SNM) in the Department of Energy Complex is vital to the national security of the United States. Active and passive nondestructive assays are used to confirm the presence of SNM in various configurations ranging from waste to nuclear weapons. Confirmation measurements for nuclear weapons are more challenging because the design complicates the detection of a distinct signal for highly enriched uranium. The emphasis of this dissertation was to investigate a new nondestructive assay technique that provides an independent and distinct signal to confirm the presence of highly enriched uranium (HEU). Once completed and tested this assay method could be applied to confirmation measurements of nuclear weapons. The new system uses a 14-MeV neutron source for interrogation and records the arrival time of neutrons between the pulses with a high efficiency detection system. The data is then analyzed by the Feynman reduced variance method. The analysis determined the amount of correlation in the data and provided a unique signature of correlated fission neutrons. Measurements of HEU spheres were conducted at Los Alamos with the new system. Then, Monte Carlo calculations were performed to verify hypothesis made about the behavior of the neutrons in the experiment. Comparisons of calculated counting rates by the Monte Carlo N-Particle Transport Code (MCNP) were made with the experimental data to confirm that the measured response reflected the desired behavior of neutron interactions in the highly enriched uranium. In addition, MCNP calculations of the delayed neutron build-up were compared with the measured data. Based on the results obtained from this dissertation, this measurement method has the potential to be expanded to include mass determinations of highly enriched uranium. Although many safeguards techniques exist for measuring special nuclear material, the number of assays that can be used to confirm HEU in shielded systems is

  4. Legal and regulatory framework of Uranium's enrichment

    International Nuclear Information System (INIS)

    Antelo, Josefina; Figueredo, Micaela S.; Mangone, Gisela P.; Manin, Maria L.; Pota, Luciana F.

    2009-01-01

    The object of this paper is to develop the legal aspects referred to the activities of uranium's enrichment, in order to achieve the pacific use of nuclear energy and to obey treatments, agreements and international conventions in which Argentine is party and through them assumes the non proliferation's commitment. In this context, we will develop the rights and obligations established in those legal instruments, as well as the juridical concerns of the eventual subscription of Argentine to the Additional Protocol approved by the Board of Governors in 1997. (author)

  5. Evaluation of economical at a uranium enrichment demonstration plant

    International Nuclear Information System (INIS)

    Sugitsue, Noritake

    2001-01-01

    In this report, the economy of technical achievement apply in the uranium enrichment demonstration plant is evaluated. From the evaluation, it can be concluded that the expected purpose was achieved because there was a definite economic prospect to commercial plant. The benefit analysis of thirteen years operation of the uranium enrichment demonstration plant also provides a financial aspect of the uranium enrichment business. Therefore, the performance, price and reliability of the centrifuge is an important factor in the uranium enrichment business. And the continuous development of a centrifuge while considering balance with the development cost is necessary for the business in the future. (author)

  6. Integrated design of SIGMA uranium enrichment plants

    International Nuclear Information System (INIS)

    Rivarola, Martin E.; Brasnarof, Daniel O.

    1999-01-01

    In the present work, we describe a preliminary analysis of the design feedbacks in a Uranium Enrichment Plant, using the SIGMA concept. Starting from the result of this analysis, a computer code has been generated, which allows finding the optimal configurations of plants, for a fixed production rate. The computer code developed includes the model of the Thermohydraulic loop of a SIGMA module. The model contains numerical calculations of the main components of the circuit. During the calculations, the main components are dimensioned, for a posterior cost compute. The program also makes an estimation of the enrichment gain of the porous membrane, for each separation stage. Once the dimensions of the main components are known, using the enrichment cascade calculation, the capital and operation costs of the plant could be determined. At this point it is simple to calculate a leveled cost of the Separative Work Unit (SWU). A numerical optimizer is also included in the program. This optimizer finds the optimal cascade configuration, for a given set of design parameters. The whole-integrated program permits to investigate in detail the feedback in the component design. Therefore, the sensibility of the more relevant parameters can be computed, with respect of the economical variables of the plant. (author)

  7. Uranium enrichment. Enrichment processes; Enrichissement de l'uranium. Procedes d'enrichissement

    Energy Technology Data Exchange (ETDEWEB)

    Alexandre, M.; Quaegebeur, J.P. [CEA Saclay, 91 - Gif sur Yvette (France)

    2009-01-15

    Despite the remarkable progresses made in the diversity and the efficiency of the different uranium enrichment processes, only two industrial processes remain today which satisfy all of enriched uranium needs: the gaseous diffusion and the centrifugation. This article describes both processes and some others still at the demonstration or at the laboratory stage of development: 1 - general considerations; 2 - gaseous diffusion: physical principles, implementation, utilisation in the world; 3 - centrifugation: principles, elementary separation factor, flows inside a centrifuge, modeling of separation efficiencies, mechanical design, types of industrial centrifuges, realisation of cascades, main characteristics of the centrifugation process; 4 - aerodynamic processes: vortex process, nozzle process; 5 - chemical exchange separation processes: Japanese ASAHI process, French CHEMEX process; 6 - laser-based processes: SILVA process, SILMO process; 7 - electromagnetic and ionic processes: mass spectrometer and calutron, ion cyclotron resonance, rotating plasmas; 8 - thermal diffusion; 9 - conclusion. (J.S.)

  8. Uranium enrichment decontamination and decommissioning fund

    International Nuclear Information System (INIS)

    1994-01-01

    One of the most challenging issues facing the Department of Energy's Office of Environmental Management is the cleanup of the three gaseous diffusion plants. In October 1992, Congress passed the Energy Policy Act of 1992 and established the Uranium Enrichment Decontamination and Decommissioning Fund to accomplish this task. This mission is being undertaken in an environmentally and financially responsible way by: devising cost-effective technical solutions; producing realistic life-cycle cost estimates, based on practical assumptions and thorough analysis; generating coherent long-term plans which are based on risk assessments, land use, and input from stakeholders; and, showing near-term progress in the cleanup of the gaseous diffusion facilities at Oak Ridge

  9. Uranium enrichment by the separation nozzle process

    International Nuclear Information System (INIS)

    Becker, E.W.; Bier, W.; Ehrfeld, W.; Schubert, K.; Schuette, R.; Seidel, D.

    1975-11-01

    The separation nozzle process for the enrichment of the light uranium isotope U-235 has been developed at the Karlsruhe Nuclear Research Center as an alternative to the gaseous diffusion and centrifuge processes. Since 1970 the STEAG company, Essen, has been involved in the commercial implementation of the nozzle process. A first separation nozzle process. A first separation nozzle demonstration plant with a separative capacity of 180 t SWU/a shall be erected in Brazil with the participation of the Brazilian company NUCLEBRAS and the German companies STEAG and INTERATOM. Methods for the mass production of separation elements were developed by industry and extensive performance tests were carried out on commercially fabricated separation elements. Two prototype separative stages were successfully tested in Karlsruhe. Besides further plant components, a prototype of a UF 6 recycle facility was developed which serves the purpose of stripping the UF 6 from the light auxiliary gas to be recycled in a separation nozzle cascade. The performance level achieved to date characterizes the separation nozzle process as reliable and feasible economically. Therefore, the erection of a separation nozzle demonstration plant can be recognized as the implementation of an enrichment process which combines a reliable and comparatively simple technology with a high potential for further improvements. (orig.) [de

  10. Optimal set of selected uranium enrichments that minimizes blending consequences

    International Nuclear Information System (INIS)

    Nachlas, J.A.; Kurstedt, H.A. Jr.; Lobber, J.S. Jr.

    1977-01-01

    Identities, quantities, and costs associated with producing a set of selected enrichments and blending them to provide fuel for existing reactors are investigated using an optimization model constructed with appropriate constraints. Selected enrichments are required for either nuclear reactor fuel standardization or potential uranium enrichment alternatives such as the gas centrifuge. Using a mixed-integer linear program, the model minimizes present worth costs for a 39-product-enrichment reference case. For four ingredients, the marginal blending cost is only 0.18% of the total direct production cost. Natural uranium is not an optimal blending ingredient. Optimal values reappear in most sets of ingredient enrichments

  11. The world market-situation for uranium and its enrichment

    International Nuclear Information System (INIS)

    Lurf, G.

    1977-01-01

    The development of the uranium market is described as well as all pertinent facts which may have contributed to the strong rise in uranium prices of the past three years. The policies of countries which may in the future become major uranium exporters are discussed. For the conversion of uranium there is sufficient capacity. However, if construction of new plants is not started soon shortages could occur in the early 80ies. The market for enrichment has characterized in past years by substantial overcapacities. If new enrichment plants are constructed according to present schedules this overcapacity may prevail into the early 90ies. (orig.) [de

  12. Determination of uranium enrichment by using gamma-spectrometric methods

    International Nuclear Information System (INIS)

    Kutnyj, D.V.; Telegin, Yu.N.; Odejchuk, N.P.; Mikhailov, V.A.; Tovkanets, V.E.

    2009-01-01

    By using commercial analysis programs MGAU (LLNL, USA) and FRAM (LANL, USA) the summary error of gamma-spectrometric uranium enrichment measurements was investigated. Uranium samples with enrichments of 0,71; 4,46 and 20,1 % were measured. The coaxial high purity germanium detector (type GC) and the planar germanium detector (type LEGe) were used as gamma-radiation detectors. It was shown that experimental equipment and mathematical software available in NSC KIPT allow us to measure uranium enrichment by nondestructive method with accuracy of not worse than 2%.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

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

  14. Development of empirical relation for isotope of uranium in enriched uranium matrix

    International Nuclear Information System (INIS)

    Srivastava, S.K.; Vidyasagar, D.; Jha, S.K.; Tripathi, R.M.

    2018-01-01

    Uranium enriched in 235 U is required in commercial light water reactors to produce a controlled nuclear reaction. Enrichment allows the 235 U isotopes to be increased from 0.71% to a range between 2% to 5% depending upon requirement. The enriched uranium in the form of sintered UO 2 pellet is used for any commercially operating boiling light water reactors. The enriched uranium fuel bundle surface swipes sample is being analysed to assess the tramp uranium as a quality control parameter. It is known that the 234 U isotope also enriched along with 235 U isotope in conventional gaseous diffusion enrichment process. The information about enrichment percentage of 234 U helps to characterize isotopic properties of enriched uranium. A few reports provide the empirical equation and graphs for finding out the specific activity, activity percentage, activity ratio of 234 U isotopes for enriched uranium. Most of them have not provided the reference for the data used and their source. An attempt has been made to model the relationship between 234 U and 235 U as a function of uranium enrichment at low level

  15. Criticality of mixtures of plutonium and high enriched uranium

    International Nuclear Information System (INIS)

    Grolleau, E.; Lein, M.; Leka, G.; Maidou, B.; Klenov, P.

    2003-01-01

    This paper presents a criticality evaluation of moderated homogeneous plutonium-uranium mixtures. The fissile media studied are homogeneous mixtures of plutonium and high enriched uranium in two chemical forms: aqueous mixtures of metal and mixtures of nitrate solutions. The enrichment of uranium considered are 93.2wt.% 235 U and 100wt.% 235 U. The 240 Pu content in plutonium varies from 0wt.% 240 Pu to 12wt.% 240 Pu. The critical parameters (radii and masses of a 20 cm water reflected sphere) are calculated with the French criticality safety package CRISTAL V0. The comparison of the calculated critical parameters as a function of the moderator-to-fuel atomic ratio shows significant ranges in which high enriched uranium systems, as well as plutonium-uranium mixtures, are more reactive than plutonium systems. (author)

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

    International Nuclear Information System (INIS)

    Hutter, J.C.; Srinivasan, B.; Vicek, M.; Vandegrift, G.F.

    1994-01-01

    Over the last several years, uranium silicide fuels have been under development as low-enriched uranium (LEU) targets for Mo-99. The use of LEU silicide is aimed at replacing the UAl x alloy in the highly-enriched uranium dissolution process. A process to recover Mo-99 from low-enriched uranium silicide is being developed at Argonne National Laboratory. The uranium silicide is dissolved in alkaline hydrogen peroxide. Experiments performed to determine the optimum dissolution procedure are discussed, and the results of dissolving a portion of a high-burnup (>40%) U 3 Si 2 miniplate are presented. Future work related to Mo-99 separation and waste disposal are also discussed

  17. Review of uranium enrichment prospects in Canada, 1976

    International Nuclear Information System (INIS)

    Developments since 1971 which affect the prospects for uranium enrichment in Canada from the federal government point of view are reviewed. The market for enriched uranium to the year 2000 is similar to that projected in 1971. The committed enrichment capacity of the world will be sufficient until 1990. The Canadian uranium mining capability may be adequate to supply an enrichment plant, but the present reserves policy along with the currently known resources are likely to restrict exports of its products during the plant life. Prices for enriched uranium produced in Canada would be higher than those reported by other proposed new plants; however, newer enrichment techniques have some potential for cost reductions. Application of enrichment with U235 (or plutonium and U233/thorium) to CANDU offers some uranium resource conservation and possible slight power cost reductions. Construction of an enrichment plant in Canada to supply the export market is less attractive in 1976 than in 1971, but there is potential for such a business in the future. (L.L.)

  18. 78 FR 75579 - Low Enriched Uranium From France

    Science.gov (United States)

    2013-12-12

    ... From France Determination On the basis of the record \\1\\ developed in the subject five-year review, the... uranium from France would be likely to lead to continuation or recurrence of material injury to an... Commission are contained in USITC Publication 4436 (December 2013), entitled Low Enriched Uranium from France...

  19. 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

  20. Development of an On-Line Uranium Enrichment Monitor

    International Nuclear Information System (INIS)

    Xuesheng, L.; Guorong, L.; Yonggang, Z.; Xueyuan, H. X.-Y.

    2015-01-01

    An on-line enrichment monitor was developed to measure the enrichment of UF6 flowing through the processing pipes in centrifuge uranium enrichment plant. A NaI(Tl) detector was used to measure the count rates of the 186 keV gamma ray emitted from 235U, and the total quantity of uranium was determined from thermodynamic characteristics of gaseous uranium hexafluoride. The results show that the maximum relative standard deviation is less than 1% when the measurement time is 120 s or more and the pressure is more than 2 kPa in the measurement chamber. There are two working models for the monitor. The monitor works normally in the continuous model, When the gas's pressure in the pipe fluctuates greatly, it can work in the intermittent model, and the measurement result is very well. The background of the monitor can be measured automatically periodically. It can control automatically electromagnetic valves open and close, so as to change the gas's quantity in the chamber. It is a kind of unattended and remote monitor, all of data can be transfer to central control room. It should be effective for nuclear materials accountability verifications and materials balance verification at uranium enrichment plant by using the monitor to monitor Uranium enrichment of gaseous uranium hexafluoride in the output end of cascade continuously. (author)

  1. Expansion of U. S. uranium enrichment capacity. Final environmental statement

    Energy Technology Data Exchange (ETDEWEB)

    1976-04-01

    Reasonably foreseeable environmental, social, economic, and technological costs and benefits of postulated expansion of U. S. enrichment capacity through the year 2000 and reasonably available alternatives to such expansion are described. Both the gas centrifuge and gaseous diffusion methods for the enrichment of uranium are considered in this impact assessment. (JGB)

  2. Uranium and nuclear. Uranium enrichment, the other challenge for the development of civil nuclear

    International Nuclear Information System (INIS)

    Meyer, Teva

    2013-01-01

    As Iran has just stated it was ready for an agreement with the West, the author proposes a discussion of issues related to uranium enrichment, an activity which is currently possible to perform in industrial quantities only in few installations located in France, Germany, Netherlands, United Kingdom, USA, Russia, China and Japan. The author outlines the existence of a new geographical dynamics of enriched uranium, and that enrichment is however a sensitive technology because of its military applications. The author then discusses three mechanisms aimed at ensuring a continuity in nuclear fuel supply: promotion of an internationalisation of means of production, creation of physical stocks of enriched uranium, and creation of supply guarantees for countries which are already producers. The author then outlines difficulties to apply these mechanisms. A map illustrates situations and levels regarding uranium enrichment production and needs

  3. The uranium enrichment industry and the SILEX process

    International Nuclear Information System (INIS)

    Goldsworthy, M.

    1999-01-01

    Silex Systems Limited has been developing a new laser isotope separation process since 1992. The principle application of the SILEX Technology is Uranium Enrichment, the key step in the production of fuel for nuclear power plants. The Uranium Enrichment industry, today worth ∼ US$3.5 Billion p.a., is dominated by four major players, the largest being USEC with almost 40% of the market. In 1996, an agreement was signed between Silex and USEC to develop SILEX Technology for potential application to Uranium Enrichment. The SILEX process is a low cost, energy efficient scheme which may provide significant commercial advantage over current technology and competing laser processes. Silex is also investigating possible application to the enrichment of Silicon, Carbon and other materials. Significant markets may develop for such materials, particularly in the semiconductor industry

  4. 77 FR 51579 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2012-08-24

    .... Complex, July 30, 2012, August Uranium (93.35%). uranium-235 high-enriched 1, 2012, XSNM3726, 11006037. contained in 7.5 uranium in the kilograms uranium. form of broken metal to the Atomic Energy of Canada... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant...

  5. Uranium enrichment (a strategy analysis overview)

    International Nuclear Information System (INIS)

    Blahnik, C.

    1979-08-01

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

  6. Material control and accounting requirements for uranium enrichment facilities

    International Nuclear Information System (INIS)

    Ting, P.

    1991-01-01

    This paper reports that the U.S. Nuclear Regulatory Commission has defined material control and accounting (MC and A) requirement for low-enriched uranium enrichment plants licensed under 10 CFR parts 40 and 70. Following detailed assessment of potential safeguards issues relevant to these facilities, a new MC and A rule was developed. The primary safeguards considerations are detection of the loss of special nuclear material, detection of clandestine production of special nuclear material of low strategic significance for unauthorized use or distribution, and detection of unauthorized production of uranium enriched to ≥10 wt % U-235. The primary safeguards concerns identified were the large absolute limit of error associated with the material balance closing, the inability to shutdown some uranium enrichment technologies to perform a cleanout inventory of the process system, and the flexibility of some of these technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could circumvent the detection of the production and removal of 5 kilograms of U-235 as high-enriched uranium through conventional material control and accounting programs. Safeguards techniques, including the use of production and process control information, measurements, and technical surveillance, were identified to compensate for these concerns

  7. 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)

  8. Transformations of highly enriched uranium into metal or oxide

    International Nuclear Information System (INIS)

    Nollet, P.; Sarrat, P.

    1964-01-01

    The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

  9. Process for enriching uranium from seawater

    International Nuclear Information System (INIS)

    Heitkamp, D.; Inden, P.

    1982-01-01

    In selective elutriation of uranium deposited on titanium oxide hydrate by carbonate solution, only uranium should be dissolved from the absorption material forming carbonate compounds, without the deposited ballast ions, above all of magnesium, calcium and sodium being elutriated. The uranium elutriation according to the invention is therefore carried out in the presence of these ballast ions in the same concentrations as those in seawater. The carbonate concentration can only be raised as far as the solubility product of the basic magnesium carbonate permits, so that magnesium remains in the solution, as well as carbonate, in the concentration present in seawater. One must accept the absence of calcium ions in the elutriation solution, as their solubility product with carbonate is considerably less than that for magnesium. (orig./PW) [de

  10. 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

  11. Uranium enrichment: investment options for the long term

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    The US government supplies a major portion of the enriched uranium used to fuel most of the nuclear power plants that furnish electricity in the free world. As manager of the US uranium enrichment concern, the Department of Energy (DOE) is investigating a number of technological choices to improve enrichment service and remain a significant world supplier. The Congress will ultimately select a strategy for federal investment in the uranium enrichment enterprise. A fundamental policy choice between possible future roles - that of the free world's main supplier of enrichment services, and that of a mainly domestic supplier - will underlie any investment decision the Congress makes. The technological choices are gaseous diffusion, gas centrifuge, and atomic vapor laser isotope separation (AVLIS). A base plan and four alternatives were examined by DOE and the Congressional Budget Office. In terms of total enterprise costs, Option IV, ultimately relying on advanced gas centrifuges for enrichment services, would offer the most economic approach, with costs over the full projection period totaling $123.5 billion. Option III, ultimately relying on AVLIS without gas centrifuge enrichment or gaseous diffusion, falls next in the sequence, with costs of $128.2 billion. Options I and II, involving combinations of the gas centrifuge and AVLIS technologies, follow closely with costs of $128.7 and $129.6 billion. The base plan has costs of $136.8 billion over the projection period. 1 figure, 22 tables

  12. Recent developments in the United States uranium enrichment enterprise

    International Nuclear Information System (INIS)

    Longenecker, J.R.

    1987-01-01

    In the near term, DOE is reducing production costs at the gaseous diffusion plants (GDPs), and we've made significant progress already. GDP production costs are expected to decline even further in the near future. DOE is also negotiating new power contracts for the GDPs. The new power contracts, capital improvements, and the use of more unfirm power should reduce our GDP average cost of production to about $60/SWU in the 1990s. Significant technical progress on the Atomic Vapor Laser Isotope Separation (AVLIS) advanced enrichment technology has been made recently. The highlight has been a series of half-scale integrated enrichment experiments using the Laser Demonstration Facility and the Mars separator. We are also ready to initiate testing in the full-scale Separator Demonstration Facility, including a 100 hour run that will vaporize over 5 tons of uranium. DOE is developing plans to restructure the enterprise into a more businesslike entity. The key objective in 1987 is to work with Congress to advance the restructuring of the U.S. uranium enrichment enterprise, to assure its long term competitiveness. We hope to establish in law the charter, objectives, and goals for the restructured enterprise. DOE expects that the world price for enrichment services will continue to decrease in the future. There should be sufficient excess enrichment capacity in the future to assure that competition will be keen. Such a healthy, competitive, world enrichment market will be beneficial to both suppliers and consumers of uranium enrichment services. (J.P.N.)

  13. Uranium enrichment : global view and Brazilian perspectives

    International Nuclear Information System (INIS)

    Zouain, D.M.; Sakamoto, L.H.

    1981-12-01

    A global view of isotope enrichment involving a general description of process (technical-economical aspects and policy) and status in developing countries is done. An enrichment demand in function of the Brazilian Nuclear Program is evaluated, analyzing a probable market and a low market. The perspectives to attend this demand, are studied. (E.G.) [pt

  14. Principles and techniques of uranium enrichment

    International Nuclear Information System (INIS)

    Frejacques, Claude; Mezin, Michel

    1975-01-01

    The main separation processes already used industrially or likely to be used before the end of century (gas diffusion, ultracentrifugation, laser, the nozzle process, a process developed in South Africa) are presented. Some data on the costs of the enrichment are clarified. The main characteristics of the enrichment market in which the Eurodif plant is called upon, on the expiration of five years, to take a foremost place are reported [fr

  15. Uranium enrichment with lasers - will South Africa lead or lag?

    International Nuclear Information System (INIS)

    Du Toit, G.

    1992-01-01

    Over 30 percent of the cost of locally made nuclear fuel in South Africa is associated with increasing the concentration of uranium 235. Cheaper enrichment technologies and, in particular, the decision by the Atomic Energy Corporation of South Africa to concentrate its research efforts on laser techniques are therefore of considerable significance. The laser isotope separation programme in South Africa is reviewed. 1 ill

  16. The jet nozzle process for uranium 235 isotopic enrichment

    International Nuclear Information System (INIS)

    Jordan, I.; Umeda, K.; Brown, A.E.P.

    1979-01-01

    A general survey of the isotopic enrichment of Uranium - 235, principally by jet nozzle process, is made. Theoretical treatment of a single stage and cascade of separation stages of the above process with its development in Germany until 1976 is presented [pt

  17. Nickel container of highly-enriched uranium bodies and sodium

    International Nuclear Information System (INIS)

    Zinn, W.H.

    1976-01-01

    A fuel element comprises highly enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel

  18. Soviet supplies of enriched uranium to capitalist countries

    International Nuclear Information System (INIS)

    Valvoda, Z.

    1977-01-01

    The Soviet supplies of enriched uranium to the following capitalist countries are surveyed: Belgium, Finland, France, FRG, Austria, Spain, Sweden, and the United Kingdom. The time period, total supplies, average annual supplies, estimated average price per separation work unit, and the date of the conclusion of the contract are reported. (J.B.)

  19. A confirmatory measurement technique for highly enriched uranium

    International Nuclear Information System (INIS)

    Sprinkle, J.K. Jr.

    1987-07-01

    This report describes a confirmatory measurement technique for measuring uranium items in their shipping containers. The measurement consists of a weight verification and the detection of three gamma rays. The weight can be determined very precisely, thus it severely constrains the options of the diverter who might want to imitate the gamma signal with a bogus item. The 185.7-keV gamma ray originates from 235 U, the 1001 keV originates from a daughter of 238 U, and the 2614 keV originates from a daughter of 232 U. These three gamma rays exhibit widely different attenuation properties, they correlate with enrichment and total uranium mass, and they rigorously discriminate against a likely diversion scenario (low-enriched uranium substitution). These four measured quantities, when combined, provide a signature that is very difficult to counterfeit

  20. 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

  1. 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

  2. DOE hands over uranium enrichment duties to government corporation

    International Nuclear Information System (INIS)

    Simpson, J.

    1993-01-01

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

  3. Enrichment technology of uranium and the proliferation problem

    International Nuclear Information System (INIS)

    Zhang Hui; Du Xiangwan

    1995-04-01

    The important way to control the proliferation of nuclear weapons was to control the production and the use of fissile material. In particular, strengthening the control on 'weapon-usable' fissile materials-highly enriched uranium (HEU) and separated plutonium is essential. The production of HEU and its proliferation problem was mainly discussed. A comprehensive review of the status of various enrichment techniques is presented and the impact of these enrichment processes on the proliferation problem is discussed by the comparison of the characteristics of these techniques. Moreover, various possibilities to produce HEU by an existing facility for the production of LEU and the processes possibly to be chosen for constructing dedicated facility in a country are also analysed. Finally, the safeguards on the enrichment technology and some possible ways to detect the clandestine enrichment process are discussed. (3 tabs., 5 figs.)

  4. Consequences of the new ICRP recommendations on uranium enrichment and uranium chemistry

    International Nuclear Information System (INIS)

    Bonnefoy-Claudet, J.

    1991-01-01

    From the first available information on the draft of new recommendations of the International Commission Radiological Protection, consequences should be very different depending upon industry type and handled products. That is to say: negligible for uranium enrichment by gaseous diffusion and important for future laser isotope separation techniques and for uranium chemistry especially for oxide treatment. This is enhanced when the products are coming from reprocessing [fr

  5. Uranium enrichment: an activity for the riches

    International Nuclear Information System (INIS)

    Mundim, S.G.

    1985-01-01

    The United States in the first place, the European countries next, have begun to maintain strict control of the technology transfer of the methods of enrichment, helped largely by the International Atomic Energy Agency which creates additional obstacles with is so-called 'safeguards'. So that it can possess a complete energy cycle, Brazil needs to have access to the techniques of enrichment. The methods are many and still extremely costly. Committed to signing a risk contract with Germany in the scope of its nuclear treaty with the intention of implementing here the jetcentrofuge method, Brazil should, nevertheless, make an effort to cause to emerge from its laboratories a new procedure which would permit it to become independent in this sector. (Author) [pt

  6. Uranium enrichment: a competitive market in the future?

    International Nuclear Information System (INIS)

    Marques, Andre Ferreira; Honaiser, Eduardo Henrique Rangel

    2005-01-01

    Uranium enrichment is the costly step in the nuclear fuel cycle. It has born as a an activity for the military in the 40s, financed by governments, such as the United States (US) and the former Soviet Union. Later, other major nations have joined them in the nuclear weapons development. The activity of enrichment was done in each country that developed nuclear weapons, and the nuclear weapons countries, especially the US and Soviet Union, dictated the mined uranium market. In the 70s, with the growth of the commercial use of nuclear energy, uranium enrichment started to be treated as a market, which gradually have structured itself, strongly influenced by the historical background. Today, the market is an oligopoly of four major government-owned (or government-influenced) companies. In this paper, the trends in the enrichment market are identified, focusing on competitiveness. Through the conduction of a market analysis (past and future), and the study of the market structure evolution, a more competitive market is shown, but still influenced by the governmental participation. Competitiveness is dictated by government support, verticalization capacity, and, mainly by technological advantages. (author)

  7. Inventory control through gamma spectrometry at the enriched uranium laboratory

    International Nuclear Information System (INIS)

    Vicens, H.E.; Korob, R.O.; Goldschmidt, A.E.

    1987-01-01

    The enriched uranium laboratory processes alternatively uranium 90% and 20% enriched in U-235. The control of the isotopic composition of lots is made through mass spectrometry. In the laboratory operation wastes of both enrichments are generated and the recovery is performed with a time delay. To strengthen the administrative controls, avoid errors related to personnel replacement and/or deferred operations, it seemed suitable to adjust the gamma spectrometry as a fast, simple and available method to determine the enrichment. The laboratory work includes a wet and a dry process. The waste recovery necessarily involves the handling of liquid samples. For this reason, it was decided to determine the calibration curve for uranyl nitrate samples of fixed concentration and geometry. The samples were prepared from material purified through double precipitation of uranium peroxide and subsequent ignition to U 3 O 8 in platinum crucible, in tubular oven during 8 hours at 720 deg C. The preparation of samples, the measurement description, the discussion of results and the analysis of errors due to the presence of insoluble material and concentration changes are included. (Author)

  8. Simulation of transportation of low enriched uranium solutions

    International Nuclear Information System (INIS)

    Hope, E.P.; Ades, M.J.

    1996-01-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes

  9. Aerodynamic isotope separation processes for uranium enrichment: process requirements

    International Nuclear Information System (INIS)

    Malling, G.F.; Von Halle, E.

    1976-01-01

    The pressing need for enriched uranium to fuel nuclear power reactors, requiring that as many as ten large uranium isotope separation plants be built during the next twenty years, has inspired an increase of interest in isotope separation processes for uranium enrichment. Aerodynamic isotope separation processes have been prominently mentioned along with the gas centrifuge process and the laser isotope separation methods as alternatives to the gaseous diffusion process, currently in use, for these future plants. Commonly included in the category of aerodynamic isotope separation processes are: (a) the separation nozzle process; (b) opposed gas jets; (c) the gas vortex; (d) the separation probes; (e) interacting molecular beams; (f) jet penetration processes; and (g) time of flight separation processes. A number of these aerodynamic isotope separation processes depend, as does the gas centrifuge process, on pressure diffusion associated with curved streamlines for the basic separation effect. Much can be deduced about the process characteristics and the economic potential of such processes from a simple and elementary process model. In particular, the benefit to be gained from a light carrier gas added to the uranium feed is clearly demonstrated. The model also illustrates the importance of transient effects in this class of processes

  10. Radioactive effluents, Portsmouth Uranium Enrichment Complex, calendar year 1983

    International Nuclear Information System (INIS)

    Acox, T.A.; Klein, L.S.

    1984-03-01

    Radioactive discharges from the Portsmouth Uranium Enrichment Complex are discussed and tabulated. Tables indicate both the location of the discharge and the nuclides discharged. Routine discharges for 1983 are well below the Radioactive Concentration Guide limits specified in DOE Order 5480.1, Chapter XI. There was, however, an unplanned release in December from the X-326 Building Side Purge which exceeded the limits. 1 figure

  11. Analysis on Enrichment and Mass of Uranium Materials With Time-Correlated Measurement.

    Science.gov (United States)

    Tian, Chenyang; Guo, Huiping; Lv, Ning; Xu, Peng; Lv, Wenhui

    2017-11-01

    Effective determination for enrichment of the uranium materials in heavy shielding is technically difficult to solve. Various technical methods so far have been proposed to deal with this problem. With the time-correlated coincidence measurement method, 14 MeV neutrons and fission spectrum neutrons are chosen to interrogate the uranium samples with different masses and enrichments. The fission spectrum neutrons are obtained by slowing down the 14 MeV neutrons. Results have shown that a good linear relationship exists between masses of the uranium materials (≤10% enriched U) that possess the same enrichment. A linear relationship also exists between enrichments of the uranium materials that possess the same mass.

  12. Critical experiments on enriched uranium graphite moderated cores

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

    1978-07-01

    A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

  13. Progress on the IPNS Enriched Uranium Booster Target

    International Nuclear Information System (INIS)

    Knox, A.E.; Carpenter, J.M.; Bailey, J.L.

    1986-09-01

    We describe the Enriched Uranium Booster Target designed for use in Argonne's Intense Pulsed Neutron Source. This report contains a general description of the system, and descriptions of the thermal-hydraulic and loss-of-coolant accident analyses, of the neutronic, criticality and power density calculations, of the assessment of radiation and thermal cycling growth, and of the disk fabrication methods. We also describe the calculations of radionuclide buildup and the related hazards analysis and our calculations of the temperature and stress profiles in the disks, and briefly allude to considerations of security and safeguards

  14. Spectrographic determination of impurities in enriched uranium solutions

    International Nuclear Information System (INIS)

    Capdevila, C.; Roca, M.

    1980-01-01

    A spectrographic procedure for the determination of trace amounts of Al, B, Ba, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, K, L i , Hg, Mn, Mo, Na, Nb, Ni, P, Pb, Ru, Sb, Sn, Sr, Ti, V, Zn, and Zr in enriched uranyl nitrate solutions from the reprocessing of spent nuclear fuels is described. After removal of uranium by either TBP or TNOA solvent extraction, the aqueous phase Is analysed by the graphite spark technique. TBP is adequate for all impurities, excepting boron and phosphorus; both of these elements can sat is factory be determined by using TNOA after the addition of mannitol to avoid boron losses. (Author) 4 refs

  15. Video monitoring system for enriched uranium casting furnaces

    Energy Technology Data Exchange (ETDEWEB)

    Turner, P.C.

    1978-03-01

    A closed-circuit television (CCTV) system was developed to upgrade the remote-viewing capability on two oralloy (highly enriched uranium) casting furnaces in the Y-12 Plant. A silicon vidicon CCTV camera with a remotely controlled lens and infrared filtering was provided to yield a good-quality video presentation of the furnace crucible as the oralloy material is heated from 25 to 1300/sup 0/C. Existing tube-type CCTV monochrome monitors were replaced with solid-state monitors to increase the system reliability.

  16. Uranium enrichment services activity. Financial statements for fiscal year ended June 30, 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Financial statements for the Uranium Enrichment Services Activity, covering both ERDA and its contractors, are presented to provide information concerning the operations and financial position of this operation

  17. Economic evaluation of the gaseous diffusion on uranium enrichment

    International Nuclear Information System (INIS)

    Naruse, Yuji; Yoshida, Hiroshi; Fujine, Sachio; Aochi, Tetsuo; Higashi, Kunio.

    1977-11-01

    A gaseous diffusion plant is a large multi-stage system, which is composed of about a thousand stages in cascade and other support facilities even to obtain the slightly enriched uranium for LWR. To design and construct such a reliable, versatile, economical plant, it is necessary to know the process characteristics and to make a techno-economic evaluation of the enrichment plant. A cascade optimization and a conceptual design of the diffusion plant are described, and the cost evaluation formulae based the 1970-1972 cost level in Japan are given. In parametric case studies with the economic evaluation program developed, technological and economic factors in the plant economy are presented. (auth.)

  18. Alpha spectrometry enriched uranium urinalysis results from IPEN

    International Nuclear Information System (INIS)

    Lima, Marina Ferreira

    2008-01-01

    Full text: IPEN (Instituto de Pesquisas Energeticas e Nucleares) manufactures the nuclear fuel to its research reactor, the IEA-R1. The CCN (Centro do Ciclo do Combustivel) facility produces the fuel cermets from UF 6 (uranium hexafluoride) enriched to 19.75% in 235 U. The production involves the transformation of the gaseous form in oxides and silicates by ceramic and metallurgical processing. The workers act in more than one step that involves exposition to types F, S and M compounds of uranium. Until 2003, only fluorimetric analysis was carried out by the LRT (Laboratorio de Radiotoxicologia - IPEN) in order to evaluate the intake of uranium, in spite of the sub estimation of the 234 U contribution to the internal doses. Isotopic uranium determination in urine by alpha spectrometry is the current method to monitoring the contribution of 234 U, 235 U and 238 U. Alpha spectrometry data of 164 samples from 84 individuals separate in three categories of workers: routinely work group; special operation group and control group - were analyzed how the isotopic composition excreted by urinary tract corresponds with the level of enrichment and isotopic composition of the plant products. Results show that is hard to estimate these intakes of 234 U and 235 U since these isotopes alpha activities are below the limit of detection or minimum detectable activity (MAD) of this method in the most part of the samples. Only in 22 samples it was possibly to measure the three radionuclides. Not expected high contribution of 234 U activity was found in samples of the control group. No one result over the 234 U and 235 U MAD was found in the samples from the special operation group. Only in 5 samples from the routinely group the levels of 235 U was higher than the levels of others groups. In a complementary study, 3 solid samples of UF 6 , U 2 O 8 and U 3 Si 2 from CCN plant were analyzed to determinate the isotopic uranium composition in these salts, since this composition varies

  19. Permit issued for expansion of the Gronau uranium enrichment plant

    International Nuclear Information System (INIS)

    2005-01-01

    Over the next few years, the Gronau Uranium Enrichment Plant (UAG) will be able to raise its capacity in steps from 1 800 tonnes to 4 500 tonnes SWU per annum. This development in steps depends on the extent to which demand on the world market can be turned into contracts for Urenco. After a procedure taking more than six years to complete, Urenco Deutschland GmbH was issued the required permit by the North Rhine-Westphalian State Ministry for Transport, Energy, and Planning on February 14 this year. The permit was preceded by the drafting and examination of two safety reports of several hundred pages each, and of over 1 000 detailed documents. The application documents were examined by a large number of expert consultants and public authorities from all over Germany acting on behalf of the nuclear licensing authority. Urenco Ltd. sells enrichment services worldwide, at present holding a share of approx. 19% of the world market for uranium separative work. Although this world market is not going to grow for a foreseeable period of time, Urenco is expanding its capacities. This is based on the latest centrifuge technology characterized by extremely low power consumption and high availability. It replaces diffusion technology, which currently holds a share of approx. 40% of the world market. (orig.)

  20. Uranium mineralization in fluorine-enriched volcanic rocks

    Energy Technology Data Exchange (ETDEWEB)

    Burt, D.M.; Sheridan, M.F.; Bikun, J.; Christiansen, E.; Correa, B.; Murphy, B.; Self, S.

    1980-09-01

    Several uranium and other lithophile element deposits are located within or adjacent to small middle to late Cenozoic, fluorine-rich rhyolitic dome complexes. Examples studied include Spor Mountain, Utah (Be-U-F), the Honeycomb Hills, Utah (Be-U), the Wah Wah Mountains, Utah (U-F), and the Black Range-Sierra Cuchillo, New Mexico (Sn-Be-W-F). The formation of these and similar deposits begins with the emplacement of a rhyolitic magma, enriched in lithophile metals and complexing fluorine, that rises to a shallow crustal level, where its roof zone may become further enriched in volatiles and the ore elements. During initial explosive volcanic activity, aprons of lithicrich tuffs are erupted around the vents. These early pyroclastic deposits commonly host the mineralization, due to their initial enrichment in the lithophile elements, their permeability, and the reactivity of their foreign lithic inclusions (particularly carbonate rocks). The pyroclastics are capped and preserved by thick topaz rhyolite domes and flows that can serve as a source of heat and of additional quantities of ore elements. Devitrification, vapor-phase crystallization, or fumarolic alteration may free the ore elements from the glassy matrix and place them in a form readily leached by percolating meteoric waters. Heat from the rhyolitic sheets drives such waters through the system, generally into and up the vents and out through the early tuffs. Secondary alteration zones (K-feldspar, sericite, silica, clays, fluorite, carbonate, and zeolites) and economic mineral concentrations may form in response to this low temperature (less than 200 C) circulation. After cooling, meteoric water continues to migrate through the system, modifying the distribution and concentration of the ore elements (especially uranium).

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

    Energy Technology Data Exchange (ETDEWEB)

    Brim, Cornelia P.

    2013-04-01

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

  2. Uranium enrichment by jet nozzle separation process in the German-Brazil cooperation program

    International Nuclear Information System (INIS)

    Becker, E.W.

    1991-01-01

    It presents a lecture on technical and commercial aspects of uranium isotopic enrichment by the Get Lozzle Method in Brazil. The analysis is presented regarding the context of bilateral agreement German-Brazil it discusses the technical problems of a demonstration plant design as well as the commercial exportation viability of enriched uranium produced in Brazil by the Jet Nozzle Method. (author)

  3. Technical problems in case of utilizing uranium of medium enrichment for a research reactor

    International Nuclear Information System (INIS)

    Kanda, Keiji; Shibata, Shun-ichi

    1979-01-01

    Usually, highly enriched uranium of 90 - 93% is used for research reactors, but the US government proposed the strong policy to use low enriched uranium of the uranium of medium enrichment in unavoidable case from the viewpoint of the resistance to nuclear proliferation in November, 1977. This policy is naturally applied to Japan also. The export of highly enriched uranium will be permitted only when the President approves it after the technical and economical evaluations by the government. The Kyoto University high flux reactor has the features which are not seen in other research reactors, such as medical irradiation, and it is hard to attain the objectives of researches unless HEU is used. The application for the export of HEU was accepted in February, 1978. The nuclear characteristics of the KUHFR when medium or low enriched uranium is used, the criticality experiment in the KUCA using the uranium of medium enrichment, and the burning test on the uranium fuel plates of medium enrichment are described. The research project to lower the degree of enrichment in the fuel for research and test reactors is expected to be continued down to less than 20%. The MEU of 45% enrichment will be actually used in 1983. (Kako, I.)

  4. Uranium and enrichment supply and demand - Facts and trends

    International Nuclear Information System (INIS)

    Stobbs, J.

    1990-01-01

    The author summarizes the general characteristics of the uranium and enrichment markets as follows: both markets have significant overcapacity; the situation has resulted from the past sales policies of producers and in part from the purchasing policies of the buyers; new low-price supply sources have appeared on the market; a continuing low level of spot prices and the perception of plentiful future supply will clearly influence buyer procurement strategies by pushing towards spot or short-term purchases or of tying long-term prices to the spot price. Data supporting the above conclusions provide a historical perspective since the mid 1970's and projections of market changes through the turn of the century

  5. Highly enriched uranium (HEU) storage and disposition program plan

    International Nuclear Information System (INIS)

    Arms, W.M.; Everitt, D.A.; O'Dell, C.L.

    1995-01-01

    Recent changes in international relations and other changes in national priorities have profoundly affected the management of weapons-usable fissile materials within the United States (US). The nuclear weapon stockpile reductions agreed to by the US and Russia have reduced the national security requirements for these fissile materials. National policies outlined by the US President seek to prevent the accumulation of nuclear weapon stockpiles of plutonium (Pu) and HEU, and to ensure that these materials are subjected to the highest standards of safety, security and international accountability. The purpose of the Highly Enriched Uranium (HEU) Storage and Disposition Program Plan is to define and establish a planned approach for storage of all HEU and disposition of surplus HEU in support of the US Department of Energy (DOE) Fissile Material Disposition Program. Elements Of this Plan, which are specific to HEU storage and disposition, include program requirements, roles and responsibilities, program activities (action plans), milestone schedules, and deliverables

  6. Validation of NCSSHP for highly enriched uranium systems containing beryllium

    International Nuclear Information System (INIS)

    Krass, A.W.; Elliott, E.P.; Tollefson, D.A.

    1994-01-01

    This document describes the validation of KENO V.a using the 27-group ENDF/B-IV cross section library for highly enriched uranium and beryllium neutronic systems, and is in accordance with ANSI/ANS-8.1-1983(R1988) requirements for calculational methods. The validation has been performed on a Hewlett Packard 9000/Series 700 Workstation at the Oak Ridge Y-12 Plant Nuclear Criticality Safety Department using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software code package. Critical experiments from LA-2203, UCRL-4975, ORNL-2201, and ORNL/ENG-2 have been identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. The results of these calculations establish the safety criteria to be employed in future calculational studies of these types of systems

  7. A 2000-2010 years outlook of isotopic uranium enrichment

    International Nuclear Information System (INIS)

    Vasaru, G.

    1998-01-01

    The increase of the installed power in nuclear plants implies the following steps to be achieved: - developing a parallel industry for the nuclear fuel cycle able to ensure a rhythmic supply of natural uranium, possibly an isotopic enrichment of 235 U of around 1.2 - 3.2%, depending on the reactor system; - manufacturing the fuel elements and the operation of cycle back-end, which may, possibly, include a temporary storage of the irradiated fuel; - reprocessing the spend fuel; - radioactive waste processing in view of final disposal, as well as the recovery of un-spent uranium and of plutonium formed. The heavy water reactors of CANDU-PHW does not imply any isotopic enrichment but provides a lower burnup of only 7,000 MW day/tone. An enrichment to 1.2% in 235 U for this type of reactors could increase the burnup up to 20,000 MW day/tone. An advanced method of enriching 235 U is based on the Atomic Vapor Laser Isotop Separation (AVLIS). This procedure called AVLIS has several advantages which are pointed out in this paper, among which: a very high selectivity; high separation factors; a low energy consumption due to the fact that in the conditions of a selective photo ionization, the energy necessary to the process is only 6.2 eV for the separated 235 U atom vs 0.3 MeV in case of inertial separators or 3 MeV in case of gaseous diffusion procedure. With the current laser yields an energy consumption of 100 kWh/SWU is estimated for AVLIS procedures as compared with 2,400 kWh/SWU in case of gaseous diffusion; an almost entire extraction of 235 U, what ensures a more efficient utilisation of nuclear fuel. Due to its modular character and to potential improvement in the equipment which could be achieved, this procedure will ensure a reduction in the investment costs in the construction stage what will make AVLIS a substitute of the classical separation procedures

  8. Uranium enrichment measurement task with a connectionist architecture

    International Nuclear Information System (INIS)

    Vigneron, V.; Martinez, J.M.; Morel, J.; Lepy, M.C.

    1995-01-01

    Layered Neural Networks, which are a class of models based on neural computation, are applied to the measurement of uranium enrichment, i.e. the isotope ration 235 U/( 235 U+ 236 U+ 238 U). The usual methods consider a limited number of γ-ray and X-ray peaks, and requires previously calibrated instrumentation for each sample. But, in practice, the source-detector ensemble geometry conditions are critically different, thus a means of improving the above conventional methods is to reduce the region of interest: this is possible by focusing on the region called K α X where the three elementary components are present. The measurement of these components in mixtures leads to the desired ratio. Real data are used to study its performance. Training is done with a Maximum Likelihood method. We show the encoding of data by Neural Networks is a promising method to measure uranium 235 U and 238 U quantities in infinitely thick samples. (authors). 7 refs., 2 figs., 1 tab

  9. Uranium enrichment measurement task with a connectionist architecture

    Energy Technology Data Exchange (ETDEWEB)

    Vigneron, V.; Martinez, J.M. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Morel, J.; Lepy, M.C. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. des Applications et de la Metrologie des Rayonnements Ionisants

    1995-12-31

    Layered Neural Networks, which are a class of models based on neural computation, are applied to the measurement of uranium enrichment, i.e. the isotope ration {sup 235} U/({sup 235} U+{sup 236} U+{sup 238} U). The usual methods consider a limited number of {gamma}-ray and X-ray peaks, and requires previously calibrated instrumentation for each sample. But, in practice, the source-detector ensemble geometry conditions are critically different, thus a means of improving the above conventional methods is to reduce the region of interest: this is possible by focusing on the region called K{sub {alpha}}X where the three elementary components are present. The measurement of these components in mixtures leads to the desired ratio. Real data are used to study its performance. Training is done with a Maximum Likelihood method. We show the encoding of data by Neural Networks is a promising method to measure uranium {sup 235} U and {sup 238} U quantities in infinitely thick samples. (authors). 7 refs., 2 figs., 1 tab.

  10. 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

  11. Material surveillance and verification program at a uranium enriching plant

    International Nuclear Information System (INIS)

    DeVito, V.J.

    1975-01-01

    A license for a nuclear facility in the United States is approved only after a licensee demonstrates by procedure or practice that an adequate material control system exists. A license can specify acceptable material control practices. Therefore, processors in the United States receiving uranium hexafluoride (UF 6 ) from a U. S. Government-owned enriching plant can accept shipper's values for nuclear material accounting purposes if: there is surveillance during withdrawal of the UF 6 , an independent sample is obtained, and certain measurement verification is subsequently performed by the receiver or the receiver's agent. Because of the high equipment and operating costs, essentially all UF 6 processors have adopted a surveillance and verification program. A resident observer is employed to perform surveillance, obtain samples, and tamper-safe the shipping cylinders. Samples are analyzed by the receiver or by an independent laboratory. The observer determines by surveillance that withdrawals, or transfers of material, weighings, and sampling are accomplished in accordance with accepted procedures. Surveillance of the withdrawals includes observing the transfer of UF 6 from the enriching plant cylinder to the shipping cylinder(s) and the withdrawal of samples. In addition, it inclu []es observing the weighing of all cylinders associated with a sample lot of UF 6 . Following the surveillance of withdrawals, weighings, and sampling, the cylinders are made tamper-safe by the application of tamper-indicating devices. Statistics for the verification program have shown shipper and receiver measurements to be within the limits acceptable for adequate material control. (auth)

  12. 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)

  13. Present-day issues of seawater uranium extraction and chemical enrichment technology

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu

    2008-01-01

    Combination of an efficient extraction process of uranium from seawater and its chemical enrichment is a promising technology having the chance to display a leadership in the global uranium market. It may serve as a role of nuclear material supply or stock thus promoting the nuclear power generation over the world. In this report a special interest is focused on the recovery of uranium from seawater by immobilized tannin with the aid of Black stream (Oceanic current) combined as a line with uranium enrichment by the ACEP (Asahi Chemical Exchange Process) process for which the separation factor of 10 -3 is reported. (S. Ohno)

  14. Status report on the cost and availability of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, Hans; Laucht, Juergen

    2005-01-01

    Availability and price development of enriched uranium contained in fuel elements for research reactors plays an important role with regard to reliability and economic and planning reasons. The leading price factors of LEU (19.75% enriched uranium metal), are the contained natural uranium equivalent in the form of UF6 (feed component), the separative work of the enrichment (SWU), conversion of the enriched uranium into metal form and associated services, such as transportation. World market price of feed material for enrichment was more or less stable in the last decades. After very moderate feed price increases between 2001 and mid-2003, the price gained momentum and almost doubled in the short period between the 2nd half of 2003 and year-end 2004. (author)

  15. Designing and analysis study of uranium enrichment with gas centrifuge

    International Nuclear Information System (INIS)

    Tsunetoshi Kai

    2006-01-01

    This note concerns a designing and analysis study of uranium enrichment with a gas centrifuge. At first, one dimensional model is presented and a conventional analytical method is applied to grasp the general idea of a centrifuge performance. Secondly, two-dimensional numerical method is adopted to describe the diffusion phenomena with assumption of simple flow patterns. Parametric surveys are made on the dimension of a centrifuge rotor, the gas feed, withdrawal and circulation system, and operation variables such as feed flow rate, cut and so on. Thirdly, full numerical solutions are obtained for the flow and diffusion equations in static state, using a modified version of the Newton method without neglect of any non-linear term. The numerical results are compared with the experimental data made by Beams et al. and Zippe, and found to be in good agreement. Further, the theoretical pressure and separative power are compared respectively with experimental ones on a comparatively recent centrifuge. The results reveal that the characteristics of separation performance of a centrifuge can be fully described by the present method. Some of inevitable problems are tackled regarding UF 6 gas isotope separation by centrifugation. To examine the influence of the extraneous light gas, the diffusion equations for ternary mixture are solved and also the flow field of binary mixture with large mass difference is obtained to simultaneously solve the Navier-Stokes equations and the diffusion equation.for binary case. Since the gas in the interior region of the rotor is so rarefied that the Navier-Stokes equations cease to be valid, the Burnett equations are solved.for gas flow in a rotating cylinder. Considering that the uranium recovered at a reprocessing plant includes 236 U besides 235 U and 238 U, the concentration distributions of the ternary gas isotopes are determined and a value function is defined for the evaluation of separative work for the multi-component mixture

  16. Energies and media nr 32. Conditions for the nuclear sector. Uranium and its enrichment for EDF. The role of Russia

    International Nuclear Information System (INIS)

    2010-07-01

    After some comments on recent events in the nuclear sector in different countries (international conference, agreement between Rosatom and the IAEA, energy policy and projects in India, Turkey, Vietnam, France, Finland, UK, Sweden, USA, Germany), this publication discusses the uranium enrichment issue for EDF and the role of Russia. It refers to a report by the French High committee for transparency and information on nuclear safety (HCTISN) and to a hearing by the French Parliament. Different aspects are discussed: the national and international legal contexts and their relationship, the EDF reactor supply (enriched uranium of MOX, fixed elements in enrichment contracts, supply security), enrichment processes and installations (in France, natural uranium enrichment, use of reprocessed or recycle uranium, use of depleted uranium as a by-product of enrichment), how EDF and AREVA are working with Russia (enrichment of recycled uranium, use of depleted uranium)

  17. Environmental Survey preliminary report, Portsmouth Uranium Enrichment Complex, Piketon, Ohio

    International Nuclear Information System (INIS)

    1987-08-01

    This report presents the preliminary findings from the first phase of the Environmental Survey of the United States Department of Energy (DOE) Portsmouth Uranium Enrichment Complex (PUEC), conducted August 4 through August 15, 1986. The Survey is being conducted by an interdisciplinary team of environmental specialists, led and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. Team specialists are being supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. This phase of the Survey involves the review of existing site environmental data, observations of the operations performed at PUEC, and interviews with site personnel. The Survey team developed a Sampling and Analysis Plan to assist in further assessing certain of the environmental problems identified during its on-site activities. The Sampling and Analysis Plan will be executed by Argonne National Laboratory. When completed, the results will be incorporated into the PUEC Environmental Survey Interim Report. The Interim Report will reflect the final determinations of the PUEC Survey. 55 refs., 22 figs., 21 tabs

  18. Environmental Survey preliminary report, Portsmouth Uranium Enrichment Complex, Piketon, Ohio

    Energy Technology Data Exchange (ETDEWEB)

    1987-08-01

    This report presents the preliminary findings from the first phase of the Environmental Survey of the United States Department of Energy (DOE) Portsmouth Uranium Enrichment Complex (PUEC), conducted August 4 through August 15, 1986. The Survey is being conducted by an interdisciplinary team of environmental specialists, led and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. Team specialists are being supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. This phase of the Survey involves the review of existing site environmental data, observations of the operations performed at PUEC, and interviews with site personnel. The Survey team developed a Sampling and Analysis Plan to assist in further assessing certain of the environmental problems identified during its on-site activities. The Sampling and Analysis Plan will be executed by Argonne National Laboratory. When completed, the results will be incorporated into the PUEC Environmental Survey Interim Report. The Interim Report will reflect the final determinations of the PUEC Survey. 55 refs., 22 figs., 21 tabs.

  19. Efficiency criteria for optimization of separation cascades for uranium enrichment

    Directory of Open Access Journals (Sweden)

    Georgy Sulaberidze

    2018-02-01

    Full Text Available As it is known, uranium enrichment is carried out on industrial scale by means of multistage separation facilities, i.e., separation cascades in which gas centrifuges (GCs are connected in series and parallel. Design and construction of these facilities require significant investment. So, the problem of calculation and optimization of cascade working parameters is still relevant today. At the same time, in many cases, the minimum unit cost of a product is related to the cascade having the smallest possible number of separation elements/GCs. Also, in theoretical studies, it is often acceptable to apply as an efficiency criterion the minimum total flow to supply cascade stages instead of the abovementioned minimum unit cost or the number of separation elements. In this article, cascades with working parameter of a single GC changing from stage to stage are optimized by two of the abovementioned performance criteria and are compared. The results obtained allow us to make a conclusion about their differences.

  20. 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)

  1. Characterization of highly enriched uranium in a nuclear forensic exercise

    International Nuclear Information System (INIS)

    Nascimento, Marcos R.L. do; Quinelato, Antonio L.; Silva, Nivaldo C. da; Sarkis, Jorge E.S.

    2011-01-01

    This paper presents the characterization of two metal samples of highly enriched uranium as a contribution of Pocos de Caldas Laboratory, LAPOC, a branch of Brazilian National Commission for Nuclear Energy, CNEN, to the Round Robin 3, R R3, coordinated by the Nuclear Forensics International Technical Working Group. A scenario was constructed in which two separate seizures of nuclear material occurred and forensics analysis was requested to help discern whether these incidents were related and whether these incidents exceeded country statutes. Laboratories were instructed to submit assessment reports in 24 hours, one week, and two month time frames. Besides preliminary evaluations for categorization of the material, our laboratory applied high resolution gamma spectrometry, optical emission spectrometry by inductively coupled plasma, and potentiometric titration for quantitative characterization of the samples. Concerning our technical reports answers for the three main forensics questions formulated by R R3, one of them was inconclusive, considering that LAPOC does not yet have all essential equipment for a fully satisfactory forensics nuclear analysis. (author)

  2. Comparative analysis of calculations and experiment for uranium-graphite lattices with natural and slightly-enriched uranium

    International Nuclear Information System (INIS)

    Khrennikov, N.N.; Shchukin, A.V.

    1988-01-01

    Three sets of experiments carried out at different times and in different laboratories on measuring the material parameter for uranium-graphite lattices using natural and slightly enriched uranium are analyzed. Comparison with the calculations by the TRIFOGR and MCU (the Monte Carlo method) codes reveals resonable agreement between the calculation and experiment (of the order of 0.4% in K eff ). 17 refs.; 3 tabs

  3. Uranium enrichment - A case study in indigenous process design and development

    International Nuclear Information System (INIS)

    Van Breda Smith, M.

    1987-01-01

    The position of indigenous process development and design is discussed in general and then with specific reference to uranium enrichment. The writer concludes, most specifically from the reported South African process design achievements in uranium enrichment, that when technology requirements for medium sized chemical projects have a strong or unique local reference, the broad and detail process design and development can and should be undertaken by indigenous engineering entities provided psychological barriers in local engineering management can be overcome

  4. Uranium enrichment - a case study in indigenous process design and development

    International Nuclear Information System (INIS)

    Smith, M.v.B.

    1987-01-01

    The position of indigenous process development and design is discussed in general and then with specific referecne to uranium enrichment. The writer concludes, most specifically from the reported South African process design achievements in uranium enrichment, that when technology requirements for medium sized chemical projects have a strong or unique local reference, the broad and detail process design and development can and should be undertaken by indigenous engineering entities provided psychological barriers in local engineering management can be overcome. 4 figs., 8 refs

  5. 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.

  6. Highly enriched uranium, a dangerous substance that should be eliminated

    International Nuclear Information System (INIS)

    Schaper, Annette

    2013-01-01

    Either highly enriched uranium (HEU) or plutonium is needed to construct a nuclear weapon. While plutonium is radioactive and hazardous in handling, HEU is far less dangerous. Furthermore, it is more difficult to detect by technical means. Therefore, in comparison to plutonium, HEU is much easier to divert, smuggle and hide. Moreover, a crude nuclear explosive made of HEU can be constructed in a much simpler way than one made using plutonium. For these reasons, HEU is the material most wanted by terrorists. A few tens of kilograms are sufficient for one explosive, but the quantities existing in the world add up to hundreds of tons. Due to the disarmament at the end of the Cold War, the NPT nuclear weapon states possess large quantities of HEU in excess of their needs for nuclear weapons. Therefore, these countries have not produced HEU for many years. Several international projects are working towards reducing the proliferation risks posed by HEU. The projects include the reduction of existing HEU by converting it to civilian reactor fuel that cannot be easily used for nuclear weapons. Other projects work towards reducing the number of countries and sites where HEU is stored by transferring it back to the countries of origin. And there are yet other projects which seek to minimize uses which would require new production of HEU. An international non-proliferation goal should be to eliminate all uses of HEU and thus to eliminate the need for any future production. Uses of HEU other than for nuclear weapons are as fuel in civilian research reactors, as base material for the production of special isotopes used in medical diagnostics, so-called medical targets and as fuel in military naval reactors. It is desirable to replace the HEU in all these applications with other materials and thus cease all HEU production forever. Use as fuel in civilian reactors has been greatly reduced during the last few decades. Within an international campaign, the Reduced Enrichment for

  7. Highly enriched uranium, a dangerous substance that should be eliminated

    Energy Technology Data Exchange (ETDEWEB)

    Schaper, Annette

    2013-07-01

    Either highly enriched uranium (HEU) or plutonium is needed to construct a nuclear weapon. While plutonium is radioactive and hazardous in handling, HEU is far less dangerous. Furthermore, it is more difficult to detect by technical means. Therefore, in comparison to plutonium, HEU is much easier to divert, smuggle and hide. Moreover, a crude nuclear explosive made of HEU can be constructed in a much simpler way than one made using plutonium. For these reasons, HEU is the material most wanted by terrorists. A few tens of kilograms are sufficient for one explosive, but the quantities existing in the world add up to hundreds of tons. Due to the disarmament at the end of the Cold War, the NPT nuclear weapon states possess large quantities of HEU in excess of their needs for nuclear weapons. Therefore, these countries have not produced HEU for many years. Several international projects are working towards reducing the proliferation risks posed by HEU. The projects include the reduction of existing HEU by converting it to civilian reactor fuel that cannot be easily used for nuclear weapons. Other projects work towards reducing the number of countries and sites where HEU is stored by transferring it back to the countries of origin. And there are yet other projects which seek to minimize uses which would require new production of HEU. An international non-proliferation goal should be to eliminate all uses of HEU and thus to eliminate the need for any future production. Uses of HEU other than for nuclear weapons are as fuel in civilian research reactors, as base material for the production of special isotopes used in medical diagnostics, so-called medical targets and as fuel in military naval reactors. It is desirable to replace the HEU in all these applications with other materials and thus cease all HEU production forever. Use as fuel in civilian reactors has been greatly reduced during the last few decades. Within an international campaign, the Reduced Enrichment for

  8. Unallocated Off-Specification Highly Enriched Uranium: Recommendations for Disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bridges, D. N.; Boeke, S. G.; Tousley, D. R.; Bickford, W.; Goergen, C.; Williams, W.; Hassler, M.; Nelson, T.; Keck, R.; Arbital, J.

    2002-02-27

    The U.S. Department of Energy (DOE) has made significant progress with regard to disposition planning for 174 metric tons (MTU) of surplus Highly Enriched Uranium (HEU). Approximately 55 MTU of this 174 MTU are ''offspec'' HEU. (''Off-spec'' signifies that the isotopic or chemical content of the material does not meet the American Society for Testing and Materials standards for commercial nuclear reactor fuel.) Approximately 33 of the 55 MTU have been allocated to off-spec commercial reactor fuel per an Interagency Agreement between DOE and the Tennessee Valley Authority (1). To determine disposition plans for the remaining {approx}22 MTU, the DOE National Nuclear Security Administration (NNSA) Office of Fissile Materials Disposition (OFMD) and the DOE Office of Environmental Management (EM) co-sponsored this technical study. This paper represents a synopsis of the formal technical report (NNSA/NN-0014). The {approx} 22 MTU of off-spec HEU inventory in this study were divided into two main groupings: one grouping with plutonium (Pu) contamination and one grouping without plutonium. This study identified and evaluated 26 potential paths for the disposition of this HEU using proven decision analysis tools. This selection process resulted in recommended and alternative disposition paths for each group of HEU. The evaluation and selection of these paths considered criteria such as technical maturity, programmatic issues, cost, schedule, and environment, safety and health compliance. The primary recommendations from the analysis are comprised of 7 different disposition paths. The study recommendations will serve as a technical basis for subsequent programmatic decisions as disposition of this HEU moves into the implementation phase.

  9. 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

  10. Competitiveness through change: institutional restructuring of the United States uranium enrichment enterprise

    International Nuclear Information System (INIS)

    Longenecker, J.R.

    1987-01-01

    The position of the United States programme of uranium enrichment under the Department of Energy is explained. Its competitiveness has improved over the past few years by normalising supply and demand and by streamlining the costs of gaseous diffusion plant production. The historical aspects of the uranium enrichment service are explained. Revised criteria to describe the guidelines to cover pricing, contracting and other crucial functions are under discussion. Two aspects of the new criteria of particular interest -restrictions on foreign-origin uranium and recovery of Government costs - are noted. Possible private sector involvement in uranium enrichment is discussed. Technological innovations are explained and equipment illustrated. These should improve the industry's competitiveness. (U.K.)

  11. 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)

  12. 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

  13. Expansion of U.S. uranium enrichment capacity. Final environmental statement

    International Nuclear Information System (INIS)

    1976-04-01

    Reasonably foreseeable environmental, social, economic, and technological costs and benefits of postulated expansion of U. S. enrichment capacity through the year 2000 and reasonably available alternatives to such expansion are described. Both the gas centrifuge and gaseous diffusion methods for the enrichment of uranium are considered in this impact assessment

  14. SIGMA, the novel approach of a new non-proliferating uranium enrichment technology

    International Nuclear Information System (INIS)

    Rivarola, M.; Florido, P.; Brasnarof, D.; Bergallo, J.

    2001-01-01

    The SIGMA concept, under development by Argentina, represents the evolution of the Uranium Enrichment Gaseous Diffusion technology, updated to face the challenge of the new economic-based and competitive world frame. The Enrichment technology has been historically considered as a highly proliferating activity in the nuclear field, and central countries have limited the access of the developing countries to this technology. The SIGMA concept incorporates innovative proliferation resistant criteria at the beginning of the design process, and inherits all the non-proliferation features of the Gaseous Diffusion Plants (GDPs). The radical new proliferation resistant approach of the SIGMA technology, suggest a new kind of global control of the Uranium Enrichment Market, were some developing countries might access to an Enrichment plant without accessing to the technology itself. In this paper, we analyse the economy of the SIGMA plants, and the implications of this technology on the Uranium Global Market. (authors)

  15. SIGMA: the novel approach of a new non-proliferating uranium enrichment technology

    International Nuclear Information System (INIS)

    Rivarola, M.; Florido, P.; Brasnarof, D.; Bergallo, E.

    2000-01-01

    The SIGMA concept, under development by Argentina, represents the evolution of the Uranium Enrichment Gaseous Diffusion technology, updated to face the challenge of the new economic-based and competitive world frame. The Enrichment technology has been historically considered as a highly proliferating activity in the nuclear field, and central countries limited the access of the developing countries to this technology. The SIGMA concept incorporates innovative proliferation resistant criteria at the beginning of the design process, and inherits all the non-proliferation features of the gaseous diffusion plants (GDPs). The radical new proliferation resistance approach of the SIGMA technology suggests a new kind of global control of the uranium enrichment market, where some developing countries might access an Enrichment plant without access to the technology itself. In this paper, we investigate the economy of the SIGMA plants, and the implications of this technology on the Uranium Global Market. (authors)

  16. Measurement of the enrichment of uranium-hexafluoride gas in product pipes in the centrifuge enrichment plant at Almelo

    International Nuclear Information System (INIS)

    Packer, T.W.; Lees, E.W.; Aaldijk, J.K.; Harry, R.J.S.

    1987-09-01

    One of the objectives of safeguarding centrifuge enrichment plants is to apply non-destructive measurements inside the cascade area to confirm that the enrichment level is in the low enriched uranium range. Research in the UK and USA has developed a NDA instrument which can confirm the presence of low enriched uranium on a rapid go/no go basis in cascade header pipework of their centrifuge enrichment plants. The instrument is based on a gamma spectroscopic measurement coupled with an X-ray fluorescence analysis. This report gives the results of measurements carried out at Almelo by the UKAEA Harwell, ECN Petten and KFA Juelich to determine if these techniques could be employed at Almelo and Gronau. The energy dispersive X-ray fluorescence analysis has been applied to determine the total mass of uranium in the gas phase, and the deposit correction technique and the two geometry technique have been applied at Almelo to correct the measured gamma intensities for those emitted by the deposit. After an executive summary the report discusses the principles of the two correction methods. A short description of the equipment precedes the presentation of the results of the measurements and the discussion. After the conclusions the report contains two appendices which contain the derivation of the formulae for the deposit correction technique and a discussion of the systematic errors of this technique. 8 figs.; 11 refs.; 6 tables

  17. Containment and storage of uranium hexafluoride at US Department of Energy uranium enrichment plants

    Energy Technology Data Exchange (ETDEWEB)

    Barlow, C.R.; Alderson, J.H.; Blue, S.C.; Boelens, R.A.; Conkel, M.E.; Dorning, R.E.; Ecklund, C.D.; Halicks, W.G.; Henson, H.M.; Newman, V.S.; Philpot, H.E.; Taylor, M.S.; Vournazos, J.P. [Oak Ridge K-25 Site, TN (United States). UEO Enrichment Technical Operations Div.; Russell, J.R. [USDOE Oak Ridge Field Office, TN (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States); Ziehlke, K.T. [MJB Technical Associates (United States)

    1992-07-01

    Isotopically depleted UF{sub 6} (uranium hexafluoride) accumulates at a rate five to ten times greater than the enriched product and is stored in steel vessels at the enrichment plant sites. There are approximately 55,000 large cylinders now in storage at Paducah, Kentucky; Portsmouth, Ohio; and Oak Ridge, Tennessee. Most of them contain a nominal 14 tons of depleted UF{sub 6}. Some of these cylinders have been in the unprotected outdoor storage environment for periods approaching 40 years. Storage experience, supplemented by limited corrosion data, suggests a service life of about 70 years under optimum conditions for the 48-in. diameter, 5/16-in.-wall pressure vessels (100 psi working pressure), using a conservative industry-established 1/4-in.-wall thickness as the service limit. In the past few years, however, factors other than atmospheric corrosion have become apparent that adversely affect the serviceability of small numbers of the storage containers and that indicate the need for a managed program to ensure maintenance ofcontainment integrity for all the cylinders in storage. The program includes periodic visual inspections of cylinders and storage yards with documentation for comparison with other inspections, a group of corrosion test programs to permit cylinder life forecasts, and identification of (and scheduling for remedial action) situations in which defects, due to handling damage or accelerated corrosion, can seriously shorten the storage life or compromise the containment integrity of individual cylinders. The program also includes rupture testing to assess the effects of certain classes of damage on overall cylinder strength, aswell as ongoing reviews of specifications, procedures, practices, and inspection results to effect improvements in handling safety, containment integrity, and storage life.

  18. Containment and storage of uranium hexafluoride at US Department of Energy uranium enrichment plants

    International Nuclear Information System (INIS)

    Barlow, C.R.; Alderson, J.H.; Blue, S.C.; Boelens, R.A.; Conkel, M.E.; Dorning, R.E.; Ecklund, C.D.; Halicks, W.G.; Henson, H.M.; Newman, V.S.; Philpot, H.E.; Taylor, M.S.; Vournazos, J.P.; Pryor, W.A.; Ziehlke, K.T.

    1992-07-01

    Isotopically depleted UF 6 (uranium hexafluoride) accumulates at a rate five to ten times greater than the enriched product and is stored in steel vessels at the enrichment plant sites. There are approximately 55,000 large cylinders now in storage at Paducah, Kentucky; Portsmouth, Ohio; and Oak Ridge, Tennessee. Most of them contain a nominal 14 tons of depleted UF 6 . Some of these cylinders have been in the unprotected outdoor storage environment for periods approaching 40 years. Storage experience, supplemented by limited corrosion data, suggests a service life of about 70 years under optimum conditions for the 48-in. diameter, 5/16-in.-wall pressure vessels (100 psi working pressure), using a conservative industry-established 1/4-in.-wall thickness as the service limit. In the past few years, however, factors other than atmospheric corrosion have become apparent that adversely affect the serviceability of small numbers of the storage containers and that indicate the need for a managed program to ensure maintenance ofcontainment integrity for all the cylinders in storage. The program includes periodic visual inspections of cylinders and storage yards with documentation for comparison with other inspections, a group of corrosion test programs to permit cylinder life forecasts, and identification of (and scheduling for remedial action) situations in which defects, due to handling damage or accelerated corrosion, can seriously shorten the storage life or compromise the containment integrity of individual cylinders. The program also includes rupture testing to assess the effects of certain classes of damage on overall cylinder strength, aswell as ongoing reviews of specifications, procedures, practices, and inspection results to effect improvements in handling safety, containment integrity, and storage life

  19. Continuing investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

    1987-01-01

    Currently much of the world's supply of 99m Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to 1) the insolubility of uranium silicides in alkaline solutions and 2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (Author)

  20. Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

    Science.gov (United States)

    McLean, W. II; Miller, P.E.; Horton, J.A.

    1995-05-02

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

  1. Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

    Science.gov (United States)

    McLean, II, William; Miller, Philip E.; Horton, James A.

    1995-01-01

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

  2. Intra-industry Affiliate Trade of Foreign Owned Companies in Poland 1993-2002

    DEFF Research Database (Denmark)

    Nielsen, Jørgen Ulff-Møller; Pawlik, Konrad

    (exports and imports) and industry variables (wages, employment, foreign capital share, investments, sales, etc.) of foreign owned companies in the Polish manufacturing industry for the years 1993-2002, this paper, investigates for the first time intra-industry affiliate trade (IIAT), including its...

  3. Domestic Multinationals and Foreign-Owned Firms in Italy: Evidence from Quantile Regression

    Directory of Open Access Journals (Sweden)

    Grasseni, Mara

    2010-06-01

    Full Text Available This paper investigates the performance differences across and within foreign-owned firms and domestic multinationals in Italy. Used for the empirical analysis are non-parametric tests based on the concept of first order stochastic dominance and quantile regression technique. The firm-level analysis distinguishes between foreign-owned firms of different nationalities and domestic MNEs according to the location of their FDI, and it focuses not only on productivity but also on differences in average wages, capital intensity, and financial and non-financial indicators, namely ROS, ROI and debt leverage. Overall, the results provide evidence of remarkable heterogeneity across and within multinationals. In particular, it seems not possible to identify a clear foreign advantage at least in terms of productivity, because foreign-owned firms do not outperform domestic multinationals. Interesting results are obtained when focusing on ROS and ROI, where the profitability gaps change as one moves from the bottom to the top of the conditional distribution. Domestic multinationals investing only in developed countries present higher ROS and ROI compared with the subgroups of foreign-owned firms, but only at the lower quantiles, while at the upper quantiles the advantage seems to favour foreign firms. Finally, in regard to domestic multinationals, there is strong evidence that those active only in less developed countries persistently exhibit the worst performances

  4. HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

    International Nuclear Information System (INIS)

    Magoulas, V; Charles Goergen, C; Ronald Oprea, R

    2008-01-01

    The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project

  5. Tetra- and hexavalent uranium forms bidentate-mononuclear complexes with particulate organic matter in a naturally uranium-enriched peatland

    DEFF Research Database (Denmark)

    Mikutta, Christian; Langner, Peggy; Bargar, John R.

    2016-01-01

    % organic C, 4.1-58.6 g/kg Fe, and up to 335 mg/kg geogenic U. Uranium was found to be heterogeneously distributed at the micrometer scale and enriched as both U(IV) and U(VI) on fibrous and woody plant debris (48 ± 10% U(IV), x̅ ± σ, n = 22). Bulk U X-ray absorption near edge structure (XANES) spectroscopy...

  6. Proposal for Monitoring Within the Centrifuge Cascades of Uranium Enrichment Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Farrar, David R.

    2017-04-01

    Safeguards are technical measures implemented by the International Atomic Energy Agency (IAEA) to independently verify that nuclear material is not diverted from peaceful purposes to weapons (IAEA, 2017a). Safeguards implemented at uranium enrichment facilities (facilities hereafter) include enrichment monitors (IAEA, 2011). Figure 1 shows a diagram of how a facility could be monitored. The use of a system for monitoring within centrifuge cascades is proposed.

  7. Open / Unipede 1995 survey on uranium and enrichment requirements and supplies in Western Europe

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    This report describes the current global situation with respect to uranium and enrichment supply and demand in Europe, as well as future prospects, based upon a survey conducted in mid-1995 by Unipede and Open among their member companies. The nuclear fuel cycle committees of Unipede and Open maintain an interest in all aspects of the nuclear fuel cycle from uranium procurement up to final disposal of radioactive waste. Regarding the front-end of the fuel cycle, the work within those committees currently concentrates of uranium and enrichment supply and demand in Europe, including the recycling of uranium and of plutonium recovered from reprocessing. For the past fifteen years, the nuclear fuel cycle committees of Unipede and Open have jointly conducted an annual survey among their members about uranium and enrichment requirements and supplies. The 1995 survey involved all nuclear electric utilities in Belgium, Finland, France, Germany, the Netherlands, Spain, Sweden, Switzerland and the United Kingdom. Exclusively electric utilities are surveyed, i.e. excluding national procurement organizations, traders, brokers, financial institutions, etc. The data obtained from the individual utilities in the same format, are aggregated and form the basis of the report. The quality of these data is guaranteed by the fact that the respondents are generally those people who are responsible for the day-to-day management of the nuclear fuel cycle in their company

  8. Nondestructive uranium enrichment determination in process holdup deposits

    International Nuclear Information System (INIS)

    Hagenauer, R.C.

    1992-01-01

    A new technique has been developed for analyzing the gamma-ray spectra of uranium compounds to determine the relative abundance of all gamma-ray emitting isotopes present. The technique works well for samples whose gamma-ray absorption properties are not well defined and has become an invaluable tool for characterizing inventory samples and residual uranium held up in process equipment. The gamma rays in a spectrum obtained with a hyperpure germanium detector are corrected for container attenuation, sample self absorption, and branching ratios. The gamma rays within a spectrum can be used to compute the container attenuation and the sample self absorption. Gamma rays emitted from deposits of uranium within a container are subject to two different absorption correction factors, container attenuation and sample self-absorption. Because of the uranium K edge effects, an X ray at 89.96 keV and a gamma ray at 143.7 keV, emitted from 235 U decay, have similar mass-attenuation properties. After correcting for the gamma-ray branching ratios, the sample self absorption correction factors for these two radiations can be held constant while the container attenuation is computed. Container thicknesses are estimated in an iterative manner until the corrected peak areas of these two radiations are the same. Once the container thickness and its corresponding correction factors are computed, then the sample self absorption correction factors can be computed using all the gamma rays in a spectrum for each particular isotope. After correcting for branching ratios, sample thicknesses are estimated and correction factors computed in an iterative manner until all the gamma and X ray peaks areas are the same for each isotope. When all the corrections are properly computed then atomic abundances are proportional to the corrected peak areas for each isotope. The technique works well on a variety of deposits

  9. Technical and economic aspects of new gaseous diffusion uranium enrichment capacity

    International Nuclear Information System (INIS)

    Langley, R.A. Jr.; O'Donnell, A.J.

    1977-01-01

    Work is well advanced on design and construction of the next major increment of U.S. uranium enrichment capacity. The plant will use the gaseous diffusion process to provide the required capacity and reliability at a competitive enrichment services cost. Gaseous diffusion technology is the base against which other processes are compared in order to assess their commercial viability. While it has generally been described as a mature technology with limited future development potential, work on design of the new U.S. plant has resulted in major improvement in plant design with corresponding decreases in plant capacity and operating costs. The paper describes major technological advances incorporated into the new plant design and their impact on enrichment costs. These include the effects of: - advanced barrier technology; - tandem compressor drive systems; - optimization of number of equipment sizes; - single level plant design; - development of rapid power level change capability; - electrical system simplification; - plant arrangement and layout. Resulting capital costs and projected enrichment costs are summarized. Enrichment costs are placed in the context of total nuclear fuel cycle costs. Trade-offs between uranium feed material quantities and enrichment plant tails assays are described, and optimization of this aspect of the nuclear fuel cycle is discussed. The effect on enrichment plant characteristics is described. Flexibility and capability of the new U.S. enrichment plant to meet these changing optimization conditions are described

  10. The technical feasibility of uranium enrichment for nuclear bomb construction at the parallel nuclear program plant

    International Nuclear Information System (INIS)

    Rosa, L.P.

    1990-01-01

    It is discussed the hole of the Parallel Nuclear Program is Brazil and the feasibility of uranium enrichment for nuclear bomb construction. This program involves two research centers, one belonging to the brazilian navy and another to the aeronautics. Some other brazilian institutes like CTA, IPEN, COPESP and CETEX and also taking part in the program. (A.C.A.S.)

  11. Air strikes on uranium enrichment plants as potential sources of radioecological danger

    Directory of Open Access Journals (Sweden)

    Živanov Dragan

    2007-01-01

    Full Text Available According to the Non Proliferation Treaty (NPT, the signatory countries are not forbidden to preform uranium enrichment for peaceful purposes. However, if there is a justified doubt that the uranium enrichment is performed with the aim to produce nuclear weapons, this certainly causes great concern. In this case, the international community can apply pressure to a certain country if it determines that the country does not want to cease activities of making its own nuclear weapons. The international community pressure on the country can be intesified until its political leadership is not made to question and cease all activities of producing nuclear weapons. This pressure can be political, economic, and as a last resort-military. As a gesture of goodwill the country can stop the uranium enrichment process. In this way, the country shows that it finally gives up the intention to produce nuclear weapons. However, when military pressure is applied, i.e. military strikes (air strikes for example on nuclear plants used for uranium enrichment, this certainly creates a risk of releasing radioactivity into the environment. That is why the aim of this paper is to signal this very fact. Using military force in these cases leads to additional radioactive contamination of the environment, so this way of solving conflicts should be avoided within the international community.

  12. 77 FR 19642 - Low Enriched Uranium From France: Final Results of Antidumping Duty Changed Circumstances Review

    Science.gov (United States)

    2012-04-02

    ... control of AREVA and the Japanese end-user. Therefore, we preliminarily determined that it was appropriate... From the Federal Register Online via the Government Publishing Office ] DEPARTMENT OF COMMERCE International Trade Administration Low Enriched Uranium From France: Final Results of Antidumping Duty Changed...

  13. 77 FR 1059 - Low Enriched Uranium From France: Initiation of Antidumping Duty Changed Circumstances Review

    Science.gov (United States)

    2012-01-09

    ... temporary shutdown of its Japanese customer caused by the earthquake and tsunami that struck Japan on March... From the Federal Register Online via the Government Publishing Office ] DEPARTMENT OF COMMERCE International Trade Administration Low Enriched Uranium From France: Initiation of Antidumping Duty Changed...

  14. Calculations of received dose for different points in the enrichment uranium oxide warehouse at 4%

    International Nuclear Information System (INIS)

    Alonso V, G.

    1990-06-01

    In order to verifying that the received dose so much inside as outside of the warehouse of enriched uranium dioxide to 4% it doesn't represent risk to the personnel, the modelling of this and the corresponding calculations for the extreme case of dose at contact are made. (Author)

  15. 78 FR 66898 - Low Enriched Uranium From France: Final Results of Changed Circumstances Review

    Science.gov (United States)

    2013-11-07

    ... DEPARTMENT OF COMMERCE International Trade Administration [A-427-818] Low Enriched Uranium From... Administration, International Trade Administration, Department of Commerce. SUMMARY: The Department of Commerce...: Andrew Huston or Mark Hoadley, AD/CVD Operations, Office VII, Enforcement and Compliance, International...

  16. A simple method for the absolute determination of uranium enrichment by high-resolution {gamma} spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Korob, R.O. [Unidad de Actividad Radioquimica y Quimica de las Radiaciones, Comision Nacional de Energia Atomica, Centro Atomico Ezeiza, Presbitero Juan Gonzalez y Aragon No. 15 Partido de Ezeiza, Provincia de Buenos Aires (Argentina)]. E-mail: korob@cae.cnea.gov.ar; Blasiyh Nuno, G.A. [Unidad de Actividad Gestion de Residuos Radiactivos, Comision Nacional de Energia Atomica, Centro Atomico Ezeiza, Presbitero Juan Gonzalez y Aragon No. 15 Partido de Ezeiza, Provincia de Buenos Aires (Argentina)

    2006-05-15

    A simple method for the determination of uranium enrichment using high-resolution {gamma} spectrometry is presented in this paper. The method relies solely on the {gamma}-ray emission probabilities of {sup 235}U and {sup 234m}Pa, and an iterative procedure for the least squares fit of a polynomial to a set of experimentally determined data. To ensure the reliability of the {sup 234m}Pa {gamma}-ray emission probabilities employed, a new determination of these probabilities was carried out using a combination of {gamma} spectrometry and Cerenkov counting of a purified {sup 234}Th solution. Using these new data, a maximum difference of {approx}5% has been found between the experimental and declared uranium enrichment in a set of solid and liquid samples containing uranium compounds.

  17. Intra-Industry Affiliate Trade of Foreign-Owned Companies in Transition Economies:

    DEFF Research Database (Denmark)

    Nielsen, Jørgen Ulff-Møller; Pawlik, Konrad

    2008-01-01

    Using a database containing trade and industry variables of foreign-owned companies in the Polish manufacturing industry for the years 1993-2002, we investigate the relationship between the organizational structure of multinational enterprises (MNE) in Poland and the intra-industry trade...... become a more important mode of organization for multinational enterprises as Poland has evolved into a relatively stable economic environment where MNEs feel comfortable basing their regional operations....

  18. Chinese business ontext in establishing of wholly foreign owned enterprise in People's Republic of China

    OpenAIRE

    Polajžer, Boštjan

    2013-01-01

    The purpose of the following article is to present research outcomes of registering a wholly foreign owned company in China, and to compare it with the establishment of a limited liability company in the European setting. With the help of an interview method, we obtained information from the selected respondents, and this information served us to verify and to compare the European and Chinese ways of company registration. The information obtained in this manner was analyzed, and served as a b...

  19. A study of the material accountancy procedure at the uranium enrichment facility

    International Nuclear Information System (INIS)

    Shirahashi, J.; Akiba, M.; Omae, M.

    1984-01-01

    This paper describes an evaluation of material accountancy based on total uranium (U element MUF) to detect diversions of significant quantity in the uranium enrichment facility operating at a stated maximum enrichment level of 5%. Verification that material production is within the declared enrichment can be achieved by the inspection activities associated with limited - frequency unannounced access (LFUA) to cascade areas as treated extensively in HSP. According to the experience of the material accountancy at our facility, the reduction of the material accountancy capability by changing from U-235 isotope MUF to U element MUF is only about half. However, still the U element MUF approach can meet the current IAEA detection goals for the up to about 1000 tswu/a plant

  20. Researches on uranium enrichment carried out in Centro de Desenvolvimento da Tecnologia Nuclear

    International Nuclear Information System (INIS)

    Campos, J.L.; Coelho, J.S.; Lemos, J.W.; Yadoya, R.; Consiglio, R.V.

    1986-01-01

    The first researches on uranium enrichment by the Jet Nozzle Process, were carried out initially by the Nuclear Research Center of Karlsruhe - KfK, in Germany. The process development received a decisive thrust in 1975, when NUCLEBRAS and other German institutions decided to join the efforts in order to demonstrate the technical and economical feasibility of the process. More recently, some experimental facilities have been transferred from KfK to the NUCLEBRAS'Research Center - CDTN, where are providing relevant data concerning to the behaviour of advanced Separation Elements and enrichment Cascades. The experience acquired by CDTN is being used to support the implantation activities of the uranium enrichment plant being erected in Resende. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

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

  2. Enriched uranium processing with 7-1/2% TBP

    International Nuclear Information System (INIS)

    Orth, D.A.; Martin, W.H.; Pickett, C.E.

    1983-01-01

    The 7-1/2% TBP flowsheet gives adequate recovery of uranium and neptunium or plutonium, with reduced waste volume as compared to the prior aluminum-salted 3-1/2% TBP flowsheet. Decontamination from fission products is sensitive to numerous variables, including aluminum nitrate concentration in the feed, impeller speeds, and prior treatment of the fuel solution in head end operations. The impeller speed in the 1A bank also influences uranium losses as well as the fission product decontamination. The magnitudes of these effects suggest that stage efficiency is poor with this flowsheet in this mixer settler unit. The existing continuous solvent washers give evidence of low washing efficiency that limits permissible feed activity and that may be related to low contact time between the solvent and the carbonate wash solution. The most general conclusion is that satisfactory operation can be obtained with all projected domestic and foreign fuels under consideration for processing, by suitable adjustment of operating conditions. Also, possible flowsheet and equipment changes are known that could improve operations with these fuels further. 7 references

  3. Application of a method to measure uranium enrichment without use of standards

    International Nuclear Information System (INIS)

    Saule, F.A.; Righetti, M.A.

    1998-01-01

    Full text: The determination of uranium enrichment in the many different stages present at a gaseous diffusion enrichment plant (diffusers, cisterns, deposits in pipes, drums with rests of process), or materials of deposit (plates of fuel elements not irradiated and recipients with uranium oxide), that have several geometries and physics properties of the containers, is very important for safeguards inspections. In this work is tested a non destructive analysis technique to determine the value of uranium enrichment of different samples with uranium materials without use of standards, to apply in safeguards inspections. It was used a hyper pure germanium detector with efficiency of 20% to obtain the gamma spectrum of the samples. In each spectrum, were used the net area values corresponding to four lines of U-235 (at 143, 163, 186 and 205 keV) and three lines of U-238 (258, 766 and 1001 keV); these values were analysed with two different methods. The comparison of the calculated and declared values showed a discrepancy of about 10%. (author) [es

  4. Assessment of liquid disposal originated by uranium enrichment at Aramar Experimental Center São Paulo--Brazil.

    Science.gov (United States)

    Gerenutti, Marli; Gonçalves, Marcos Moisés; Rissato, Sandra Regina; de Oliveira, José Martins; dos Santos Reigota, Marco Antonio; Galhiane, Mário Sergio

    2012-07-01

    This work presents a liquid disposal monitoring originated from uranium enrichment process at Aramar Experimental Center from 1990 to 1998. Assessment of uranium, fluorides, ammoniacal nitrogen, chemical oxygen demand, and pH measurements were made in water samples and compared with results achieved in other countries, as North America and India. The liquid disposal evaluation, generated by uranium enrichment process, showed low levels, considering most parameters established by Federal and State Legislation, aiming environmental pollution control. However, uranium levels were above the limits established by Conselho Nacional do Meio Ambiente, Environment Protection Agency and mainly by the World Health Organization.

  5. Initial report on characterization of excess highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

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

  6. The enterprise-wide risk management system of CNNC Lanzhou Uranium Enrichment Co., Ltd

    International Nuclear Information System (INIS)

    Liu Yinliang; Li Xuehong; Zhao Zhifu

    2010-01-01

    This paper targeted at the enterprise-wide risk management system of CNNC Lanzhou uranium enrichment co., Ltd., combining the Central enterprise-wide risk management guidelines document, outlining the main content of the overall risk management system on target, structure, process, strategy, internal control, information and culture. And it provides specific measures in evaluation and monitoring to enhance the total risk management system of CNNC Lanzhou uranium co, ltd on the basis of following the principles of total risk management system. and provides proper insurance for the overall target of risk management. (authors)

  7. Japan-IAEA sefeguards demonstration programme in the gas centrifuge uranium enrichment facility

    International Nuclear Information System (INIS)

    Akiba, Mitsunori; Iwamoto, Tomonori; Omae, Masayoshi

    1985-01-01

    The Hexa-partite Safequard Project was started for the purpose of examining the effective techniques of safeguards for gas centrifuge uranium enrichment facilities. By the proposal of respective participating countries, it was decided to carry out the verifying test of various safeguard techniques at the actual plants. Japan carried out the verifying test of safeguard techniques at the Ningyotoge uranium enrichment pilot plant in June, 1982, and from November, 1983, to August, 1984. The contents of this test is reported. In Japan, this verifying test was positioned as a part of JASPAS (Japanese project of supporting IAEA safeguards). The verifying test of realtime and in-operation inventories, the verifying test of IAEA load cell type weighing machines for UF 6 cylinders, the verifying test of the measurement of the degree of enrichment in UF 6 cylinders by nondestructive test, the verifying test of confinement/watch system, and the verifying test of IAEA gas phase uranium enrichment monitors were carried out. The results were presented as the data for examination in the HSP, and evaluated as useful, informative and well compiled. It is necessary to pursue more cost-effective approaches. (Kako, I.)

  8. The separation nozzle process for uranium isotope enrichment

    International Nuclear Information System (INIS)

    Becker, E.W.

    1977-01-01

    The paper covers the most important steps in the technological development and the future prospects of the separation nozzle process. In this process uranium isotope separation is brought about by the mass dependence of the centrifugal forces in a curved flow of a UF 6 /H 2 mixture. Due to the large excess in hydrogen, the high ratio of UF 6 flow velocity to thermal velocity required for an effective isotope separation is obtained at relatively low expansion ratios and, accordingly, with relatively low gas-dynamic losses. As the optimum Reynolds number of the curved jet is comparatively low, and as a high absolute pressure is essential for economic reasons, the characteristic dimensions of the nozzle systems are made as small as possible. For commercial application in the near future, systems involving mechanical jet deflection have been developed. Promising results were, however, also obtained with separation nozzle systems generating a streamline curvature by the interaction of opposed jets. Most of the development work has been done at the Nuclear Research Centre, Karlsruhe. Since 1970 the STEAG company (FRG) has been involved in the commercial implementation of the process. Two industrial-scale separative stages were tested successfully. This work constitutes the basis of planning of a separation nozzle demonstration plant to be built in Brazil. (author)

  9. Development and technical implementation of the separation nozzle process for enrichment of uranium 235

    International Nuclear Information System (INIS)

    Syllus Martins Pinto, C.; Voelcker, H.; Becker, E.W.

    1977-12-01

    The separation nozzle process for the enrichment of uranium-235 has been developed at the Karlsruhe Nuclear Research Center as an alternative to the gaseous diffusion and centrifuge process. The separation of uranium isotopes is achieved by the deflection of a jet of uranium hexafluoride mixed with hydrogen. Since 1970, the German company of STEAG, has been involved in the technological development and commercial implementation of the nozzle process. In 1975, the Brazilian company of NUCLEBRAS, and the German company of Interatom, joined the effort. The primary objective of the common activity is the construction of a separation nozzle demonstration plant with an annual capacity of about 200 000 SWU and the development of components of a commercial plant. The paper covers the most important steps in the development and the technical implementation of the process. (orig.) [de

  10. 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

  11. 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.

  12. Research and economic evaluation on uranium enrichment by gaseous diffusion process in Japan

    International Nuclear Information System (INIS)

    Aochi, T.; Takahashi, S.

    1977-01-01

    Research and development works on uranium enrichment by gaseous diffusion process were carried out by JAERI, IPCR and industries since 1965. There are two important keys to reduce the uranium separation cost. One is the characteristics of the barrier and the other is financing and/or political planning. The technics to prepare the barrier with pore diameter of 40A have been developed with polytetrafluoroethylene, alumina and nickel. The experiment on corrosion behavior of PTFE barriers has shown better characteristics than the others. In the field of engineering research, the adiabatic efficiency of axial compressor for UF 6 was resulted to as high as 90% by long term operation tests. Based on these experimental data, techno-economic evaluation on a uranium enrichment plant was carried out with regard to the optimization of separation efficiency, numbers of step and operating conditions of the plant. Sensitivity in the separation cost were calculated as a function of pore diameter, uranium hexafluoride cost, plant capacity, electric power cost, and the plant annual expenditure. A financing plan must be such as to achieve 1. maximization of debt in a percentage of total capitalization 2. off-take contracts to utilities as security for financing 3. minimization of risks to equity and achievable cost of capital. Therefore the cash flow analysis and the schedule for construction and operation are very important for a economical feasibility of a uranium enrichment plant. To minimize the risk, not only economical but also political environment are important. The governmental supports and international agreements will be necessary

  13. Research on and economic evaluation of uranium enrichment by gaseous diffusion in Japan

    International Nuclear Information System (INIS)

    Aochi, T.; Takahashi, S.

    1977-01-01

    Research and development on uranium enrichment by the gaseous diffusion process have been carried out by the Japan Atomic Energy Research Institute, the Institute of Physical and Chemical Research, and industries since 1965. The paper describes the two important keys to reducing the cost of uranium separation. One is the characteristics of barriers and the other is financing and/or political planning. The techniques of preparing a barrier with pore diameter 40A have been developed with polytetrafluoroethylene (PTFE), alumina and nickel. Experiments on corrosion behaviour have indicated that PTFE barriers are the most favourable. In the field of engineering research, the adiabatic efficiency of the axial compressor for UF 6 was raised to as high as 90% by long-term operation tests. Based on these experimental data, technico-economic evaluation of a uranium enrichment plant was carried out for optimization of separation efficiency, number of steps and plant operating conditions. Sensitivity in the separation cost was calculated as a function of pore diameter, cost of uranium hexafluoride, plant capacity, cost of electric power, and annual expenditure of the plant. A finance plan must be such as to achieve: (a) maximization of debt in a percentage of total capital; (b) off-take contracts to utilities as security for financing; (c) minimization of risks to equity and achievable cost of capital. Therefore, the cash flow analysis and the schedule for construction and operation are very important for the economic feasibility of a uranium enrichment plant. To minimize the risk, the economic as well as the political environment is important. Government support and international agreements are necessary. (author)

  14. 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)

  15. Safeguards considerations for uranium enrichment facilities, as applied to gas centrifuge and gaseous diffusion facilities

    International Nuclear Information System (INIS)

    1979-03-01

    The goals and objectives of IAEA safeguards as they are understood by the authors based on published documents are reviewed. These goals are then used to derive safeguards concerns, diversion strategies, and potential safeguards measures for four base cases, the production of highly enriched uranium (HEU) at a diffusion plant, the diversion of low enriched uranium (LEU) at a diffusion plant, the diversion of HEU at a gas centrifuge plant, and the diversion of LEU at a gas centrifuge plant. Tables of estimated capabilities are given for each case, under the assumption that the inspector would have access: to the cascade perimeter at or after the start of operations, to the cascade perimeter throughout construction and operation, to the cascade perimeter during operation plus a one-time access to the cascade itself, to the cascade during construction but only its perimeter during operation, or to the cascade itself during construction and operation

  16. Critical experiments in AQUILON with fuels slightly enriched in uranium 235 or in plutonium

    International Nuclear Information System (INIS)

    Chabrillac, M.; Ledanois, G.; Lourme, P.; Naudet, R.

    1964-01-01

    Reactivity comparisons have been, made in Aquilon II between geometrically identical lattices differing only by the composition of the fuel. The fuel elements consist in metallic uranium single rods with either slight differences of the isotopic composition (0.69 - 0.71 - 0.83 - 0.86 per cent of uranium 235) or slight additions of plutonium (0.043 per cent). Five lattices pitches have bean used, in order to produce a large variation of spectrum. Two additional sets of plutonium fuels are prepared to be used in the same conditions. The double comparisons: natural enriched 235 versus natural-enriched plutonium are made in such a way that a very precise interpretation is permitted. The results are perfectly consistent which seems to prove that the calculation methods are convenient. Further it can been inferred that the usual data, namely for the ratio of the η of 235 U and 239 Pu seem reliable. (authors) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. Preliminary investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Chaiko, D.J.; Heinrich, R.R.; Kucera, E.T.; Jensen, K.J.; Poa, D.S.; Varma, R.; Vissers, D.R.

    1986-11-01

    This paper presents the results of preliminary studies on the effects of substituting low enriched uranium (LEU) for highly enriched uranium (HEU) in targets for the production of fission product 99 Mo. Issues that were addressed are: (1) purity and yield of the 99 Mo//sup 99m/Tc product, (2) fabrication of LEU targets and related concerns, and (3) radioactive waste. Laboratory experimentation was part of the efforts for issues (1) and (2); thus far, radioactive waste disposal has only been addressed in a paper study. Although the reported results are still preliminary, there is reason to be optimistic about the feasibility of utilizing LEU targets for 99 Mo production. 37 refs., 1 fig., 5 tabs

  1. Use of Low Enriched Uranium Fuel in Accelerator Driven Subcritical Systems

    International Nuclear Information System (INIS)

    2017-08-01

    This publication presents the results and conclusions of an international research collaboration devoted to gaining a better understanding of the physics of Accelerator Driven Subcritical Systems (ADS), with particular emphasis on using low enriched uranium (LEU) fuel. The publication contains information on nine ADS facilities, including descriptions of the hardware deployed, experiments conducted, computational resources and procedures used in the analyses, principal results obtained, and conclusions drawn from the knowledge gained as a consequence of this work. It is intended to provide information for users of ADS systems and those involved in the design of new ADS facilities to use LEU fuel and in the conversion of some existing facilities from using highly enriched Uranium (HEU) to LEU.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  3. New approach for safeguarding enriched uranium hexafluoride bulk transfers

    International Nuclear Information System (INIS)

    Doeher, L.W.; Pontius, P.E.; Whetstone, J.R.

    1978-01-01

    The unique concepts of American National Standard ANSI N15.18-1975 ''Mass Calibration Techniques for Nuclear Material Control'' are discussed in regard to the establishment and maintenance of control of mass measurement of Uranium Hexafluoride (UF 6 ) both within and between facilities. Emphasis is placed on the role of control of the measurements between facilities, and thus establish decision points for detection of measurement problems and making safeguards judgments. The unique concepts include the use of artifacts of UF 6 packaging cylinders, calibrated by a central authority, to introduce the mass unit into all of the industries' weighing processes. These are called Replicate Mass Standards (RMS). This feat is accomplished by comparing the RMS to each facility's In-House Standards (IHS), also artifacts, and thence the usage of these IHS to quantify the systematic and random errors of each UF 6 mass measurement process. A recent demonstration, which exchanged UF 6 cylinders between two facilities, who used ANSI N15.18-1975 concepts and procedures is discussed. The discussion includes methodology and treatment of data for use in detection of measurement and safeguards problems. The discussion incorporates the methodology for data treatment and judgments concerning (1) the common base, (2) measurement process off-sets, (3) measurement process precision, and (4) shipper-receiver bulk measurement differences. From the evidence gained in the demonstration, conclusions are reached as to the usefulness of the realistic criteria for detection of mass measurement problems upon acceptance of the concepts of ANSI N15.18-1975

  4. Defects and improvement measures on emergency work of Lanzhou uranium enrichment company

    International Nuclear Information System (INIS)

    Li Haobo; Tao Lijuan; Yan Hongjuan; Xia Huiqiang

    2014-01-01

    The emergency work status of Lanzhou Uranium Enrichment company is discussed in this article. In terms of the analysis of the emergency organization.emergency plan system, emergency drills.emergency personnel responding capacity and other aspects, the auther points out some defects of the current emergency work and proposed some improvement measures, in order to further enhance the emergency response capacity of this company and to ensure the safe and stable operation of nuclear facilities. (authors)

  5. Re-investment, Survival and the Embeddedness of Foreign-Owned Plants

    OpenAIRE

    Colin Wren; Jonathan Jones

    2003-01-01

    "3,000 Jobs to go in Motorola Closure", "Black and Decker cuts 1,000 jobs", "Vauxhall axes 2,000", "North-east jobs go as Viasystems fails", "NEC to close Scottish plant", "1,900 jobs to go as Ford confirms closure". These headlines are all taken from a national newspaper over a recent few years and they indicate an important phenomenon: the much-observed closure of large-scale foreign-owned plants, often in manufacturing. In response to these high failure rates the economic development agenc...

  6. Research reactor preparations for the air shipment of highly enriched uranium from Romania

    International Nuclear Information System (INIS)

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

    2010-01-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation (RF) for conversion to low enriched uranium (LEU). The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR-S research reactor at Magurele, Romania, to Ozersk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation for Atomic Energy Rosatom and the International Atomic Energy Agency (IAEA). Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel. (author)

  7. Measurement of the enrichment of uranium in the pipework of a gas centrifuge enrichment plant

    International Nuclear Information System (INIS)

    Packer, T.W.; Lees, E.W.; Close, D.; Nixon, K.V.; Pratt, J.C.; Strittmatter, R.

    1985-01-01

    The US and UK have been separately working on the development of a NDA instrument to determine the enrichment of gaseous UF 6 at low pressures in cascade header pipework in line with the conclusions of the Hexapartite Safeguards Project viz. the instrument is capable of making a ''go/no go'' decision of whether the enrichment is less than/greater than 20%. Recently, there has been a series of very useful technical exchanges of ideas and information between the two countries. This has led to a technical formulation for such an instrumentation based on γ-ray spectrometry which, although plant-specific in certain features, nevertheless is based on the same physical principles. Experimental results from commercially operating enrichment plants are very encouraging and indicate that a complete measurement including set up time on the pipe should be attainable in about 30 minutes when measuring pipes of diameter around 110 mm. 5 refs., 4 figs

  8. 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

  9. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2009-12-10

    The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL{reg_sign} processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 {le} H/Fissile {le} 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k{sub eff}) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k{sub eff} is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately {+-} 0.001. For the cases where the reported benchmark k{sub eff} was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k{sub eff} is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k{sub eff} limit for calculations of the intermediate enriched uranium type systems.

  10. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    2009-01-01

    The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL(reg s ign) processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 (le) H/Fissile (le) 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k eff ) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k eff is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately ± 0.001. For the cases where the reported benchmark k eff was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k eff is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k eff limit for calculations of the intermediate enriched uranium type systems.

  11. HRM in foreign-owned agricultural and food processing companies in Serbia

    Directory of Open Access Journals (Sweden)

    Ratković Tatjana

    2015-01-01

    Full Text Available This paper was aimed at analysing the degree to which foreign-owned agricultural and food processing companies in Serbia implement key human resource management (HRM processes, determining whether these processes are performed in-house or outsourced, and identifying the variables which influence the decision on performing these processes within the company or outsourcing. The analysis of data collected through a survey conducted in 42 companies has confirmed the expectations that the surveyed companies mainly perform most of HRM activities The conclusion drawn from the study is that all examined HRM activities are mainly performed in-house, rather than outsourced, while the mode of performing key HRM processes mainly depends on the size of the company and the number of employees in HR department, implying that foreign-owned companies operating in agricultural and food processing sector in this transition country need to take these variables into consideration when deciding whether to perform crucial HRM activities by themselves or outsource.

  12. Measure of uranium enrichment by 14 MeV neutron irradiation

    International Nuclear Information System (INIS)

    Rezende, H.R.

    1987-01-01

    A non-destructive technique for the determination of uranium in UO 2 samples was developed, marking use of the change in the fission cross of a nuclide with the neutron energy. The active interrogation method was used by irradiating the samples with pulsed 14 MeV neutrons and furtherdetection of delayed fission neutrons. In order to descriminated U-238 from U-235 the neutron energy was tailored by means of two concentric cylinders of lead and paraffin/poliethylene, 11 and 4 cm thick. Between neutron pulses, delayed neutrons from fission were detected by a long counter built with five BF 3 proportional counters. Calibration curves for enrichment and total mass versus delayed neutron response were obtained using available UO 2 pellets of Known enrichment. Enrichment detection limit, obtained with 95% confidence level by the the Student distribution was estimated to be 0.33%. The minimal detectable mass was estimated to be 4.4 g. (Author) [pt

  13. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    Energy Technology Data Exchange (ETDEWEB)

    Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  14. Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

  15. Management of high enriched uranium for peaceful purposes: Status and trends

    International Nuclear Information System (INIS)

    2005-06-01

    Arms control agreements between some Nuclear Weapon States have led to the dismantling of many of the nuclear weapons in their military stockpiles, which in turn have produced stockpiles of excess weapons-grade high enriched uranium (HEU) from the dismantled weapons. Considering the proliferation potential of HEU, the management, control and disposition of this fissile material has become a primary focus of nuclear non-proliferation efforts worldwide. To lessen the proliferation threat of excess HEU stockpiles, the USA agreed to purchase several tonnes of excess Russian HEU down-blended to low enriched uranium (LEU). Proliferation concerns about HEU have also resulted in a global effort to convert research reactors from HEU to LEU fuel and to minimize civilian use of HEU. This publication addresses HEU management declared excesses, non-proliferation programmes and options for the use of HEU stockpiles, including disposition programmes. Also addressed are the influence of LEU derived from surplus HEU on the global market for uranium, technical issues associated with utilization and the disposition of HEU

  16. Experimental critical parameters of enriched uranium solution in annular tank geometries

    Energy Technology Data Exchange (ETDEWEB)

    Rothe, R.E.

    1996-04-01

    A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant`s Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all.

  17. 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

  18. Experimental critical parameters of enriched uranium solution in annular tank geometries

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1996-04-01

    A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant's Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all

  19. Development of dissolution process for metal foil target containing low enriched uranium

    International Nuclear Information System (INIS)

    Srinivasan, B.; Hutter, J.C.; Johnson, G.K.; Vandegrift, G.F.

    1994-01-01

    About six times more low enriched uranium (LEU) metal is needed to produce the same quantity of 99 Mo as from a high enriched uranium (HEU) oxide target, under similar conditions of neutron irradiation. In view of this, the post-irradiation processing procedures of the LEU target are likely to be different from the Cintichem process procedures now in use for the HEU target. The authors have begun a systematic study to develop modified procedures for LEU target dissolution and 99 Mo separation. The dissolution studies include determination of the dissolution rate, chemical state of uranium in the solution, and the heat evolved in the dissolution reaction. From these results the authors conclude that a mixture of nitric and sulfuric acid is a suitable dissolver solution, albeit at higher concentration of nitric acid than in use for the HEU targets. Also, the dissolver vessel now in use for HEU targets is inadequate for the LEU target, since higher temperature and higher pressure will be encountered in the dissolution of LEU targets. The desire is to keep the modifications to the Cintichem process to a minimum, so that the switch from HEU to LEU can be achieved easily

  20. Low-enriched uranium high-density target project. Compendium report

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, George; Brown, M. Alex; Jerden, James L.; Gelis, Artem V.; Stepinski, Dominique C.; Wiedmeyer, Stanley; Youker, Amanda; Hebden, Andrew; Solbrekken, G; Allen, C; Robertson., D; El-Gizawy, Sherif; Govindarajan, Srisharan; Hoyer, Annemarie; Makarewicz, Philip; Harris, Jacob; Graybill, Brian; Gunn, Andy; Berlin, James; Bryan, Chris; Sherman, Steven; Hobbs, Randy; Griffin, F. P.; Chandler, David; Hurt, C. J.; Williams, Paul; Creasy, John; Tjader, Barak; McFall, Danielle; Longmire, Hollie

    2016-09-01

    At present, most 99Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassembly of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.

  1. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  2. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  4. Validation of the Monte Carlo Criticality Program KENO V. a for highly-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Knight, J.R.

    1984-11-01

    A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results.

  5. Validity of Hansen-Roach cross sections in low-enriched uranium systems

    International Nuclear Information System (INIS)

    Busch, R.D.; O'Dell, R.D.

    1991-01-01

    Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the standard for use in k eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ p , for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ p to characterize resonance self shielding. Three prescriptions for calculating σ p are given. Finally, results of several calculations of k eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems. (Author)

  6. Validity of Hansen-Roach cross sections in low-enriched uranium systems

    International Nuclear Information System (INIS)

    Busch, R.D.; O'Dell, R.D.

    1991-01-01

    Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the ''standard'' for use in k eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ p , for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ p to characterize resonance self shielding. Three prescriptions for calculating σ p are given. Finally, results of several calculations of k eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems

  7. Environmental monitoring program of the uranium enrichment facility Almirante Alvaro Alberto

    International Nuclear Information System (INIS)

    Hiromoto, G.; Jacomino, V.M.F.; Venturini, L.; Moreira, S.R.D.; Gordon, A.M.P.L.; Duarte, C.L.; Pocequilo, B.R.S.; Mazzilli, B.P.

    1988-11-01

    In this report, the Environmental Monitoring Program of the Uranium Enrichment Facility Almirante Alvaro Alberto is outlined and the results obtained during the preoperational period are presented. Information concerning the population distribution, the use of water and land, the local agricultural production and the local meteorology are also available. In order to evaluate the levels of the background radiation, sample of water, air and biological and terrestrial indicators were analysed. Measurements were performed of natural gamma emitters concentrations levels and of uranium in air, surface water, precipitation, groundwater, river sediment, soil, grass, vegetation and various foodstuffs. For direct measurement of background radiation levels a solid state dosimeter network was used. Results are also presented for the analysis of non radioactive pollutants in the water samples and for the particles and gaseous fluorides concentration in the atmosphere. (author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-30

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

  9. Development of Industrial-Scale Fission 99Mo Production Process Using Low Enriched Uranium Target

    Directory of Open Access Journals (Sweden)

    Seung-Kon Lee

    2016-06-01

    Full Text Available Molybdenum-99 (99Mo is the most important isotope because its daughter isotope, technetium-99m (99mTc, has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of 99Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of 99Mo technology developments. Most of the industrial-scale 99Mo processes have been based on the fission of 235U. Recently, important issues have been raised for the conversion of fission 99Mo targets from highly enriched uranium to low enriched uranium (LEU. The development of new LEU targets with higher density was requested to compensate for the loss of 99Mo yield, caused by a significant reduction of 235U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission 99Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the 99Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

  10. Development of industrial-scale fission {sup 99}Mo production process using low enriched uranium target

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Kon; Lee, Jun Sig [Radioisotope Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Beyer, Gerd J. [Grunicke Strasse 15, Leipzig (Germany)

    2016-06-15

    Molybdenum-99 ({sup 99}Mo) is the most important isotope because its daughter isotope, technetium-99m ({sup 99}mTc), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of {sup 99}Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of {sup 99}Mo technology developments. Most of the industrial-scale {sup 99}Mo processes have been based on the fission of {sup 235}U. Recently, important issues have been raised for the conversion of fission {sup 99}Mo targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of {sup 99}Mo yield, caused by a significant reduction of {sup 235}U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission {sup 99}Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the {sup 99}Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

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

    International Nuclear Information System (INIS)

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-01

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

  13. Pollution and wet cleaning of separation nozzle systems for enrichment of uranium-235

    International Nuclear Information System (INIS)

    Bacher, W.; Bier, W.; Linder, N.

    1980-06-01

    Operational defects in separation nozzle plants resulting in air leaking into the system may cause permanent pollution of the narrow slits of the separation elements by products of the hydrolysis of UF 6 . The deposits may deteriorate the separation performance of the separation elements to such an extent that their further use for uranium enrichment is no longer feasible. Tests performed on commercial-scale separation element tubes indicated that the deposits can be removed by a wet chemical process effectively enough to restore the full separative power of the elements. The aspects of the technical application of the cleanup process are discussed. (orig.) [de

  14. Current status of uranium enrichment by way of chemical exchange reactions

    International Nuclear Information System (INIS)

    El Basyouny, A.; Bechthold, H.C.; Knoechel, A.; Vollmer, H.J.

    1985-04-01

    For this report, conference proceedings, patents and other types of literature have been collected to present an account of the current status of uranium enrichment by way of chemical exchange reactions. The report further presents a new concept along with the relevant process strategy developed by the authors. The principal process of the new concept is a chemical exchange process with crown ethers, complexed or free, playing an important part in the reactions. The authors also describe their experiments carried out for establishing suitable chemical systems. (orig./PW) [de

  15. Method for estimate the economic characteristics of an uranium enrichment plant by gaseous diffusion

    International Nuclear Information System (INIS)

    Berault, J.C.

    1975-01-01

    To estimate the economic characteristics of an uranium enrichment plant by gaseous diffusion is to determine the prospective price of the separative work unit to which leads the concerned technology, and to collect the data allowing to ascertain that this price remains in the area of development of the prices forecasted by the other projects. The prospective price estimated by the promoter is the synthesis of the components of the go decision and which are a potential market and a comprehensive industrially proven plant design, including the basic economic and technical data of the project. Procedures for estimating these components and their synthesis, exclusive of financing problems are reviewed [fr

  16. Disposition of highly enriched uranium obtained from the Republic of Kazakhstan. Environmental assessment

    International Nuclear Information System (INIS)

    1995-05-01

    This EA assesses the potential environmental impacts associated with DOE's proposal to transport 600 kg of Kazakhstand-origin HEU from Y-12 to a blending site (B ampersand W Lynchburg or NFS Erwin), transport low-enriched UF6 blending stock from a gaseous diffusion plant to GE Wilmington and U oxide blending stock to the blending site, blending the HEU and uranium oxide blending stock to produce LEU in the form of uranyl nitrate, and transport the uranyl nitrate from the blending site to USEC Portsmouth

  17. Development of a computer systems for operational data acquisition of uranium isotopic enrichment pilot plant

    International Nuclear Information System (INIS)

    Maia, W.M.C.

    1985-01-01

    A pilot plant for uranium enrichment using the jet nozzle process was transfered from Federal Republic of Germany to Brazil, to train Brazilian technicist in its operation and to improve the process. This pilot plant is monitored by a data acquisition system and the possibility of faulty events would cause serious dificulties, as far as maintenance is concerned (for instance, unvailable special components). It is described the development of a new system, which is proposed in order to minimize difficulties with maintenance that utilizes in the assembling integrated circuits of large scale of integration. It is controlled by a microcomputer. (Author) [pt

  18. Pion-Muon Concentrating System for Detectors of Highly Enriched Uranium

    CERN Document Server

    Kurennoy, Sergey; Blind, Barbara; Jason, Andrew J; Neri, Filippo

    2005-01-01

    One of many possible applications of low-energy antiprotons collected in a Penning trap can be a portable muon source. Released antiprotons annihilate on impact with normal matter producing on average about 3 charged pions per antiproton, which in turn decay into muons. Existence of such negative-muon sources of sufficient intensity would bring into play, for example, detectors of highly enriched uranium based on muonic X-rays. We explore options of collecting and focusing pions and resulting muons to enhance the muon flux toward the detector. Simulations with MARS and MAFIA are used to choose the target material and parameters of the magnetic system consisting of a few solenoids.

  19. Comparison Of A Neutron Kinetics Parameter For A Polyethylene Moderated Highly Enriched Uranium System

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie, IV, George Espy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, Joetta Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Grove, Travis Justin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-04-17

    This paper examines the comparison of MCNP® code’s capability to calculate kinetics parameters effectively for a thermal system containing highly enriched uranium (HEU). The Rossi-α parameter was chosen for this examination because it is relatively easy to measure as well as easy to calculate using MCNP®’s kopts card. The Rossi-α also incorporates many other parameters of interest in nuclear kinetics most of which are more difficult to precisely measure. The comparison looks at two different nuclear data libraries for comparison to the experimental data. These libraries are ENDF/BVI (.66c) and ENDF/BVII (.80c).

  20. Disposition of highly enriched uranium obtained from the Republic of Kazakhstan. Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-05-01

    This EA assesses the potential environmental impacts associated with DOE`s proposal to transport 600 kg of Kazakhstand-origin HEU from Y-12 to a blending site (B&W Lynchburg or NFS Erwin), transport low-enriched UF6 blending stock from a gaseous diffusion plant to GE Wilmington and U oxide blending stock to the blending site, blending the HEU and uranium oxide blending stock to produce LEU in the form of uranyl nitrate, and transport the uranyl nitrate from the blending site to USEC Portsmouth.

  1. Analysis of civilian processing 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. The analysis addresses several options in reducing the excess separated plutonium and HEU, and the consequences on nonproliferation and safeguards policy assessments resulting from the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials

  2. Nonproliferation analysis of the 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. The analysis addresses several options in reducing the excess separated plutonium and HEU, and the consequences on nonproliferation and safeguards policy assessments resulting from the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials

  3. Achievements of the French laser uranium enrichment program and final demonstrations

    International Nuclear Information System (INIS)

    Bordier, G.

    2006-01-01

    The research on Atomic Vapor Laser Isotope Separation (SILVA in French) started in France during the seventies at the Commissariat a l'Energie Atomique (CEA) through basic experiments on uranium spectroscopy and evaporation. The program gained momentum rapidly with the decision in 1985 to focus on SILVA while decreasing the research on chemical exchange process, centrifugation and laser molecular route. Similar orientations were taken during the same period in the US and in Japan. The French need in uranium enrichment service, around 5 MSWU (million Separation Work Units) for 45 electronuclear reactors in 1985, was fulfilled by gaseous diffusion in EURODIF (multinational consortium). But the expected increase in the international need, especially in Asia, and the high cost of the energy power supply necessary for gaseous diffusion were the main reasons to search for a very efficient and selective new generation enrichment process like SILVA. The scientific demonstration of the SILVA process was completed in the mid 1990's with grams of separated uranium at both the requested industrial enriched and tail assays simultaneously. However, in the late 90's, some technical difficulties appeared in the scaling up of the process to produce kilograms of enriched uranium. In 1998 a scientific peer review defined some general goals and improvements to be performed, partly reached on the Pelleas facility in Saclay, two years later. In parallel the centrifugation technology made important progress (in Europe, URENCO consortium and in Japan) related to the emergence of high performance carbon fibers. In September 2000, a French ministerial nuclear committee chose the centrifugation as the reference industrial technology to substitute for gaseous diffusion in 2010-2015: SILVA was not enough mature and competitive to be developed at an industrial scale within such a delay. In order to save properly the knowledge and to account for potential future uses, the committee settled on a

  4. Analysis of organizational options for the uranium enrichment enterprise in relation to asset divesture

    International Nuclear Information System (INIS)

    Harrer, B.J.; Hattrup, M.P.; Dase, J.E.; Nicholls, A.K.

    1986-08-01

    This report presents a comparison of the characteristics of some prominent examples of independent government corporations and agencies with respect to the Department of Energy's (DOE) uranium enrichment enterprise. The six examples studied were: the Bonneville Power Administration (BPA); the Tennessee Valley Authority (TVA); the Synthetic Fuels Corporation (SYNFUELS); the Consolidated Rail Corporation (CONRAIL); the British Telecommunications Corporation (British TELECOM); and the Communications Satellite Organization (COMSAT), in order of decreasing levels of government ownership and control. They range from BPA, which is organized as an agency within DOE, to COMSAT, which is privately owned and free from almost all regulations common to government agencies. Differences in the degree of government involvement in these corporations and in many other characteristics serve to illustrate that there are no accepted standards for defining the characteristics of government corporations. Thus, historical precedent indicates considerable flexibility would be available in the development of enabling legislation to reorganize the enrichment enterprise as a government corporation or independent government agency

  5. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  6. Fuel handling accident analysis for the University of Missouri Research Reactor's High Enriched Uranium to Low Enriched Uranium fuel conversion initiative

    Science.gov (United States)

    Rickman, Benjamin

    In accordance with the 1986 amendment concerning licenses for research and test reactors, the MU Research Reactor (MURR) is planning to convert from using High-Enriched Uranium (HEU) fuel to the use of Low-Enriched Uranium (LEU) fuel. Since the approval of a new LEU fuel that could meet the MURR's performance demands, the next phase of action for the fuel conversion process is to create a new Safety Analysis Report (SAR) with respect to the LEU fuel. A component of the SAR includes the Maximum Hypothetical Accident (MHA) and accidents that qualify under the class of Fuel Handling Accidents (FHA). In this work, the dose to occupational staff at the MURR is calculated for the FHAs. The radionuclide inventory for the proposed LEU fuel was calculated using the ORIGEN2 point-depletion code linked to the MURR neutron spectrum. The MURR spectrum was generated from a Monte Carlo Neutron transPort (MCNP) simulation. The coupling of these codes create MONTEBURNS, a time-dependent burnup code. The release fraction from each FHA within this analysis was established by the methodology of the 2006 HEU SAR, which was accepted by the NRC. The actual dose methodology was not recorded in the HEU SAR, so a conservative path was chosen. In compliance to NUREG 1537, when new methodology is used in a HEU to LEU analysis, it is necessary to re-evaluate the HEU accident. The Total Effective Dose Equivalent (TEDE) values were calculated in addition to the whole body dose and thyroid dose to operation personnel. The LEU FHA occupational TEDE dose was 349 mrem which is under the NRC regulatory occupational dose limit of 5 rem TEDE, and under the LEU MHA limit of 403 mrem. The re-evaluated HEU FHA occupational TEDE dose was 235 mrem, which is above the HEU MHA TEDE dose of 132 mrem. Since the new methodology produces a dose that is larger than the HEU MHA, we can safely assume that it is more conservative than the previous, unspecified dose.

  7. SRTC criticality safety technical review: Nuclear Criticality Safety Evaluation 93-04 enriched uranium receipt

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Review of NMP-NCS-930087, open-quotes Nuclear Criticality Safety Evaluation 93-04 Enriched Uranium Receipt (U), July 30, 1993, close quotes was requested of SRTC (Savannah River Technology Center) Applied Physics Group. The NCSE is a criticality assessment to determine the mass limit for Engineered Low Level Trench (ELLT) waste uranium burial. The intent is to bury uranium in pits that would be separated by a specified amount of undisturbed soil. The scope of the technical review, documented in this report, consisted of (1) an independent check of the methods and models employed, (2) independent HRXN/KENO-V.a calculations of alternate configurations, (3) application of ANSI/ANS 8.1, and (4) verification of WSRC Nuclear Criticality Safety Manual procedures. The NCSE under review concludes that a 500 gram limit per burial position is acceptable to ensure the burial site remains in a critically safe configuration for all normal and single credible abnormal conditions. This reviewer agrees with that conclusion

  8. 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.

  9. Results and prospects of radiotoxicological monitoring of workers exposed to natural or enriched uranium compounds

    International Nuclear Information System (INIS)

    Chalabreysse, J.

    1981-05-01

    The workers under supervision are exposed to natural or enriched uranium compounds in highly, moderately or slightly transferable form. Apart from systematic uranium determinations in the urine, occasional analyses of the stools and chest examinations, the working premises are also monitored. The urinary uranium test, an indirect method of measuring internal contamination, is considered of great value in the routine supervision of workers exposed to uraniferous compounds. It has brought out several important aspects, especially that of having a more accurate idea of the elimination time for the various compounds at the factory stage: for transferable compounds this is seldom longer than 48h. Another point is that the new classification introducing the notion of moderately transferable compounds (class w), of biological elimination period about 50 days, is found to be much more realistic than the old one where this notion was not accounted for (UO 3 , UF 4 ). Finally in the light of this test the authors question whether the biological elimination period of little transferable compounds is as long as might be expected (500d), and intend to examine this idea in more detail [fr

  10. 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.

  11. A study of industrial exposure to uranium aerosols from the laser enrichment procedure - methods and results

    International Nuclear Information System (INIS)

    Ansoborlo, E.; Claraz, M.; Henge-Napoli, M.H.; Metivier, H.

    1995-01-01

    Comprehensive studies of the radiotoxicological risk at new uranium enrichment processing facilities using laser isotopic separation, were particularly motivated by the generation of a uranium oxide aerosol identified as UO 2 + U metal . Taking the new ICRP 66 recommendations into account, the following study on this uranium oxide mixture, was aimed at determining the physico-chemical and biokinetic specific parameters required in order to calculate the effective dose. The activity median aerodynamic diameters (AMAD) ranged between 5.2 and 10 μm with, in some cases, up to 20% of submicron size particles, while concentration values at the workplace ranged from 1.8 to 125 Bq m -3 and biological half-time calculations gave a 48 d period with in vitro dissolution test and a 77 d period with in vivo inhalation experiments. Transfer rates and dissolution rates obtained from both in vitro and vivo experiments intend to emphasize a class W behaviour in term of ICRP 30 and M in term of ICRP 66. (authors). 3 figs., 4 tabs., 22 refs

  12. Using low-enriched uranium in research reactors: The RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1994-01-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of 99 Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program

  13. Uranium enrichment conference on modified contract option, Oak Ridge, Tennessee, July 22, 1975

    International Nuclear Information System (INIS)

    1975-01-01

    The questions and answers presented in this document originated at an industry-wide meeting sponsored by the Energy Research and Development Administration held in Oak Ridge, Tennessee, on July 22, 1975, to discuss features and provisions of an ERDA plan to adjust contracts held by firms receiving uranium enriching services from ERDA. On June 19, 1975, ERDA announced terms of an expanded contract modification plan. The modified contract option broadened a previous plan proposed on January 15, 1975, by the former Atomic Energy Commission. The meeting in Oak Ridge on July 22, 1975, was designed to provide additional information on the expanded contract option and to offer ample opportunity for questions and answers prior to August 18, 1975, by which time enriching services customers who chose the one-time option had to so notify ERDA. The meeting included presentations by officials of ERDA Headquarters and ERDA's Oak Ridge Operations on the features of the contract adjustment offer, including provisions for contract termination in whole, separative work schedule adjustments, and uranium feed delivery schedule relaxation. (auth)

  14. Packaging and transportation of derived enriched uranium for the ''megatons to megawatts'' USA/Russia agreement

    International Nuclear Information System (INIS)

    Darrough, E.; Ewing, L.; Ravenscroft, N.

    1998-01-01

    In January 1998 the United States Enrichment Corporation (USEC) and Techsnabexport Co., Ltd (TENEX) of Russia celebrated the fourth anniversary of the signing of the 20-year contract between these two executive agents. USEC and TENEX are responsible for implementing the Government to-Government agreement between the United States and the Russian Federation for the purchase of uranium derived from dismantled nuclear weapons from the former Soviet Union. This program, entitled 'Megatons to Megawatts', is the first time nuclear warheads have been turned into fuel as well as the first time a commercial contract has been used to implement such a program. As of the fourth anniversary, the equivalent of almost 1,200 nuclear warheads had been converted to fuel. USEC is responsible for making all of the arrangements to transport the Russian LEU derived from HEU--hence the term, derived enriched uranium (DEU)--from St Petersburg. Russia to the USEC plant near portsmouth, Ohio. Edlow International Company is working with USEC to implement the shipping campaign and is responsible for coordination of the port delivery within Russia, as well. The organization responsible for these shipments within Russia is IZOTOP. While the program has been a major new responsibility for USEC, the early years of the program prepared all parties for the future challenges such as increased numbers of shipments, additional originating sites in Russia and witnessing requirements in Russia. (authors)

  15. Office of Environmental Management Uranium Enrichment Decontamination and Decommissioning Fund financial statements, September 30, 1995 and 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-21

    The Energy Policy Act of 1992 (Act) requires the Department of Energy to retain ownership and responsibility for the costs of environmental cleanup resulting from the Government`s operation of the three gaseous diffusion facilities located at the K-25 site in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. The Act transferred the uranium enrichment enterprise to the United States Enrichment Corporation (USEC) as of July 1, 1993, and established the Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) to: Pay for the costs of decontamination and decommissioning at the diffusion facilities; pay the annual costs for remedial action at the diffusion facilities to the extent that the amount in the Fund is sufficient; and reimburse uranium/thorium licensees for the costs of decontamination, decommissioning, reclamation, and other remedial actions which are incident to sales to the Government.

  16. Enriched-uranium feed costs for the High-Temperature Gas-Cooled reactor: trends and comparison with other reactor concepts

    International Nuclear Information System (INIS)

    Thomas, W.E.

    1976-04-01

    This report discusses each of the components that affect the unit cost for enriched uranium; that is, ore costs, U 3 O 8 to UF 6 conversion cost, costs for enriching services, and changes in transaction tails assay. Historical trends and announced changes are included. Unit costs for highly enriched uranium (93.15 percent 235 U) and for low-enrichment uranium (3.0, 3.2, and 3.5 percent 235 U) are displayed as a function of changes in the above components and compared. It is demonstrated that the trends in these cost components will probably result in significantly less cost increase for highly enriched uranium than for low-enrichment uranium--hence favoring the High-Temperature Gas-Cooled Reactor

  17. Investigation of the fire at the Uranium Enrichment Laboratory. Analysis of samples and pressurization experiment/analysis of container

    International Nuclear Information System (INIS)

    Akabori, Mitsuo; Minato, Kazuo; Watanabe, Kazuo

    1998-05-01

    To investigate the cause of the fire at the Uranium Enrichment Laboratory of the Tokai Research Establishment on November 20, 1997, samples of uranium metal waste and scattered residues were analyzed. At the same time the container lid that had been blown off was closely inspected, and the pressurization effects of the container were tested and analyzed. It was found that 1) the uranium metal waste mainly consisted of uranium metal, carbides and oxides, whose relative amounts were dependent on the particle size, 2) the uranium metal waste hydrolyzed to produce combustible gases such as methane and hydrogen, and 3) the lid of the outer container could be blown off by an explosive rise of the inner pressure caused by combustion of inflammable gas mixture. (author)

  18. A data base for PHW reactor operating on a once-through, low enriched uranium-thorium cycle

    International Nuclear Information System (INIS)

    Lungu, S.

    1984-04-01

    The study of a detailed data base for a new once-through uranium-thorium cycle using low enriched uranium (4 and 5,5% wt. U-235) and distinct UO 2 and ThO 2 fuel channels has been performed. With reference to a standard 638 MWe CANDU-type PHWR with 380 channels, evaluation of economics, fuel behaviour and safety has been performed. The Feinberg-Galanin method (code FEINGAL) has been used for calculation of axial flux distribution. All parameters have been provided by LATREP code following up the irradiation history. Economical assessment has shown that this fuel cycle is competitive with the natural uranium fuel cycle for 1979-based values of the parameters. Fuel behaviour and safety features modelling has shown that core behaviour of the uranium-thorium reactor under abnormal and accident conditions would be at least as good as that of the standard natural uranium reactor

  19. U.S. non-proliferation policy and programs regarding use of high-enriched uranium in research reactors

    International Nuclear Information System (INIS)

    Lewis, Richard A.

    1993-01-01

    Uranium enriched to 90-93%, supplied by the U.S., is now used in 141 research and test reactors in 35 countries around the world with a cumulative power of 1714 MW. Since of the order of 3 kg of 235-U is involved annually in fuel fabrication, fresh fuel transport and storage, reactor operation, and spent fuel cooling and return per megawatt of research reactor power, it is estimated that more than 5000 kg of very high-enriched uranium is handled each year to operate these reactors. Recent U.S. assessments have led to the tentative conclusion that in only approximately 11 of these reactors, generally those of highest power or power density, is the use of 90-93% enriched uranium currently a technical necessity. Universal use of the best state-of-the-art fuel technology would permit an estimated 90 of these reactors to use 20% enriched fuel, and estimated 40 others to use 45% enriched fuel, without significant performance degradation. If advanced research reactor fuel development programs currently under way in the U.S. and elsewhere are successful, it may, in fact, be possible to operate virtually all of these reactors on less than 20% enriched uranium in the longer term. The physical and economic practicality of these developmental fuels must, of course, await future assessments. The wide-spread use of 90-93% enriched uranium today is a result of fuel technology limitations prevailing 15 years ago, the ready availability of uranium of this enrichment from the U.S. in the past, and the costs involved in introduction of new fuel technologies and enrichments into existing fuel fabrication facilities and reactors. Because of the limited reactor fuel market and the substantial costs involved in gaining safety and operating approvals for changes to reactor designs there is great pressure on fuel fabricators and reactor operators to standardize on well-proven fuel designs, and not to adopt marginal technological improvements--particularly in the absence of apparent

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Nuclear Engineering Division; Racah Inst. of Physics

    2010-09-30

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

  2. Performance and economic penalties of some LEU [low enriched uranium] conversion options for the Australian Reactor HIFAR

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Robinson, G.S.

    1987-01-01

    Performance calculations for the conversion of HIFAR to low enriched uranium (LEU) fuel have been extended to a wide range of 235 U loadings per fuel element. Using a simple approximate algorithm for the likely costs of LEU compared with highly enriched uranium (HEU) fuel elements, the increases in annual fuelling costs for LEU compared with HEU fuel are examined for a range of conversion options involving different performance penalties. No significant operational/safety problems were found for any of the options canvassed. (Author)

  3. 22 CFR 126.6 - Foreign-owned military aircraft and naval vessels, and the Foreign Military Sales program.

    Science.gov (United States)

    2010-04-01

    ... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Foreign-owned military aircraft and naval vessels, and the Foreign Military Sales program. 126.6 Section 126.6 Foreign Relations DEPARTMENT OF STATE... Industrial Security Program Operating Manual, or (7) For transfers of defense services: (i) A contract or...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

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

  5. RADIO FREQUENCY IDENTIFICATION DEVICES: EFFECTIVENESS IN IMPROVING SAFEGUARDS AT GAS-CENTRIFUGE URANIUM-ENRICHMENT PLANTS

    International Nuclear Information System (INIS)

    JOE, J.

    2007-01-01

    Recent advances in radio frequency identification devices (RFIDs) have engendered a growing interest among international safeguards experts. Potentially, RFIDs could reduce inspection work, viz. the number of inspections, number of samples, and duration of the visits, and thus improve the efficiency and effectiveness of international safeguards. This study systematically examined the applications of RFIDs for IAEA safeguards at large gas-centrifuge enrichment plants (GCEPs). These analyses are expected to help identify the requirements and desirable properties for RFIDs, to provide insights into which vulnerabilities matter most, and help formulate the required assurance tests. This work, specifically assesses the application of RFIDs for the ''Option 4'' safeguards approach, proposed by Bruce Moran, U. S. Nuclear Regulatory Commission (NRC), for large gas-centrifuge uranium-enrichment plants. The features of ''Option 4'' safeguards include placing RFIDs on all feed, product and tails (F/P/T) cylinders, along with WID readers in all FP/T stations and accountability scales. Other features of Moran's ''Option 4'' are Mailbox declarations, monitoring of load-cell-based weighing systems at the F/P/T stations and accountability scales, and continuous enrichment monitors. Relevant diversion paths were explored to evaluate how RFIDs improve the efficiency and effectiveness of safeguards. Additionally, the analysis addresses the use of RFIDs in conjunction with video monitoring and neutron detectors in a perimeter-monitoring approach to show that RFIDs can help to detect unidentified cylinders

  6. Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

    Directory of Open Access Journals (Sweden)

    Baschwitz Anne

    2016-01-01

    Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

  7. Calculation and analysis for a series of enriched uranium bare sphere critical assemblies

    International Nuclear Information System (INIS)

    Yang Shunhai

    1994-12-01

    The imported reactor fuel assembly MARIA program system is adapted to CYBER 825 computer in China Institute of Atomic Energy, and extensively used for a series of enriched uranium bare sphere critical assemblies. The MARIA auxiliary program of resonance modification MA is designed for taking account of the effects of resonance fission and absorption on calculated results. By which, the multigroup constants in the library attached to MARIA program are revised based on the U.S. Evaluated Nuclear Data File ENDF/B-IV, the related nuclear data files are replaced. And then, the reactor geometry buckling and multiplication factor are given in output tapes. The accuracy of calculated results is comparable with those of Monte Carlo and Sn method, and the agreement with experiment result is in 1%. (5 refs., 4 figs., 3 tabs.)

  8. Short-term versus long-term contracting for uranium enrichment services

    International Nuclear Information System (INIS)

    Rudy, G.P.

    1990-01-01

    The US Department of Energy (US DOE) is the world's largest and most experienced supplier of uranium enrichment services. Through the late 1970s and early 1980s, emerging market forces transformed what was once a monopoly into a highly competitive industry. In the early 1980's the DOE lost market share. But as we enter the 1990s, new market forces have emerged. The US DOE believes a responsible balance between long-term and short-term contracting will be the key to success and the key to assuring the long-term health and reliability of the nuclear fuel industry. The US DOE intends to be in this nuclear business for a long time and will continue to offer reliable and responsive services second to none

  9. Financial audit: DOE Uranium Enrichment Activity financial statements - September 30, 1984

    International Nuclear Information System (INIS)

    1986-01-01

    In GAO's opinion the Department of Energy did not comply with the terms and provisions of laws and regulations for the financial transactions tested. Specifically, DOE did not comply with the Atomic Energy Act of 1954, as amended, requiring the Uranium Enrichment Activity to recover the cost of its operations when setting prices to be charged to its customers. Also, DOE's actions concerning its new utility services contract were not in compliance with applicable statutory principles. UEA did not amend its pricing criteria as required by the Atomic Energy Act, as amended. Consequently UEA did not comply with the 45-day notification to the appropriate congressional oversight committees as also required by the act

  10. 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.

  11. Production of Low Enriched Uranium Nitride Kernels for TRISO Particle Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    McMurray, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Silva, C. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Helmreich, G. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, T. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, J. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, J. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, R. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lindemer, T. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-01

    A large batch of UN microspheres to be used as kernels for TRISO particle fuel was produced using carbothermic reduction and nitriding of a sol-gel feedstock bearing tailored amounts of low-enriched uranium (LEU) oxide and carbon. The process parameters, established in a previous study, produced phasepure NaCl structure UN with dissolved C on the N sublattice. The composition, calculated by refinement of the lattice parameter from X-ray diffraction, was determined to be UC0.27N0.73. The final accepted product weighed 197.4 g. The microspheres had an average diameter of 797±1.35 μm and a composite mean theoretical density of 89.9±0.5% for a solid solution of UC and UN with the same atomic ratio; both values are reported with their corresponding calculated standard error.

  12. 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.

  13. The feasibility of uranium enrichment in Brazil for use in nuclear bombs and the conceptual project of a nuclear explosive

    International Nuclear Information System (INIS)

    Rosa, L.P.

    1990-05-01

    This work reports the steps to define a brazilian system of nuclear safeguards under the congress responsibility. It discusses as well the feasibility of uranium enrichment for nuclear weapons, the construction of a nuclear submarine and the conceptual project of a nuclear explosive. (A.C.A.S.)

  14. 10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

    Science.gov (United States)

    2010-01-01

    ... more in the isotope U235 (for centrifuge enrichment facilities this requirement does not apply to each... for: (i) Performing, unless otherwise required to satisfy part 75 of this chapter, a dynamic... dynamic physical inventory of in-process uranium and U235 so as to provide a total plant material balance...

  15. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Science.gov (United States)

    2010-01-01

    ... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

  16. Disposition of Highly Enriched Uranium (HEU) and Pu from nuclear weapons

    International Nuclear Information System (INIS)

    Neff, T.L.

    1992-01-01

    Last year, as the Soviet Union began to crumble and the Bush-Gorbachev talks had advanced to consider the dismantling of actual warheads instead of mere delivery systems, Dr. Thomas L. Neff, a senior member of MIT's Center for International Studies, tinkered with the notion that one day soon the US could be buying Russian uranium from scrapped nuclear arms. He also considered the costly business of dismantlement and set to work on an ingenious proposal. The basic idea was simply to purchase the HEU from Russia using funds derived from savings in the US Department of Energy's enrichment enterprise. The proposal, now part of an umbrella agreement between the US and Russia announced in early September, promises large strategic benefits for the US in terms of both financing dismantlement (to the tune of $100 million annually), as well as political oversight for the operation itself. In the words of Dr. Neff, who made briefings to both governments on the proposal, open-quotes It's a budget-neutral, win-win solution.close quotes What follows is an illustrated, step-by-step analysis of the proposal, as well as a reprint of Dr. Neff's paper, Disposition of HEU and Pu from Nuclear Weapons, as presented to the Uranium Institute's annual symposium last month

  17. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  18. Enrichment of specific protozoan populations during in situ bioremediation of uranium-contaminated groundwater

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Dawn; Giloteaux, L.; Williams, Kenneth H.; Wrighton, Kelly C.; Wilkins, Michael J.; Thompson, Courtney A.; Roper, Thomas J.; Long, Philip E.; Lovley, Derek

    2013-07-28

    The importance of bacteria in the anaerobic bioremediation of groundwater polluted with organic and/or metal contaminants is well-recognized and in some instances so well understood that modeling of the in situ metabolic activity of the relevant subsurface microorganisms in response to changes in subsurface geochemistry is feasible. However, a potentially significant factor influencing bacterial growth and activity in the subsurface that has not been adequately addressed is protozoan predation of the microorganisms responsible for bioremediation. In field experiments at a uranium-contaminated aquifer located in Rifle, CO, acetate amendments initially promoted the growth of metal-reducing Geobacter species followed by the growth of sulfate-reducers, as previously observed. Analysis of 18S rRNA gene sequences revealed a broad diversity of sequences closely related to known bacteriovorous protozoa in the groundwater prior to the addition of acetate. The bloom of Geobacter species was accompanied by a specific enrichment of sequences most closely related to the amoeboid flagellate, Breviata anathema, which at their peak accounted for over 80% of the sequences recovered. The abundance of Geobacter species declined following the rapid emergence of B. anathema. The subsequent growth of sulfate-reducing Peptococcaceae was accompanied by another specific enrichment of protozoa, but with sequences most similar to diplomonadid flagellates from the family Hexamitidae, which accounted for up to 100% of the sequences recovered during this phase of the bioremediation. These results suggest a prey-predator response with specific protozoa responding to increased availability of preferred prey bacteria. Thus, quantifying the influence of protozoan predation on the growth, activity, and composition of the subsurface bacterial community is essential for predictive modeling of in situ uranium bioremediation strategies.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  20. 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

  1. Operational experience in the production of 131Molybdenum and 99Iodine with high and low uranium enrichment

    International Nuclear Information System (INIS)

    Bravo, C.; Cristini, Pablo R..; Novello, A.; Bronca, M.; Cestau, Daniel; Centurion, R.; Bavaro, R.; Cestau, J.; Gualda, E.; Bronca, P.; Carranza, Eduardo C.

    2009-01-01

    In 1992, in an effort to curtail use of Highly Enriched Uranium (HEU), hoping to alleviate nuclear security concerns, United States passed the Schumer amendment to the Energy Policy Act. This legislation conditioned U.S. export of HEU to foreign companies, understanding that these companies would switch as soon as possible to Lowly Enriched Uranium (LEU). This paper describes 99 Mo production flow chart, characteristics of process cells, shielding, systems of manipulation at distance, cell ventilation system and the method for personal dose monitoring. Production evolution for the span of years 1998 to 2007 is given by indicators, keeping in mind enrichment proportion change. Evolution shown on the indicators is directly related to the application of Safety Culture concepts adopted by personnel. (author)

  2. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Developments in the the uranium industry in Australia that took place during the quarter ended 30 June 1980 are reviewed. These include uranium mine production and uranium exploration. Prices for uranium oxide and uranium hexafluoride as at the end of June 1980 and figures for U 3 O 8 production and export from 1978 to March 1980 are listed

  3. Laboratory demonstration of a multi-sensor unattended cylinder verification station for uranium enrichment plant safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Goodman, David I [Univ. of Michigan, Ann Arbor, MI (United States); Rowland, Kelly L [Univ. of California, Berkeley, CA (United States); Smith, Sheriden [Colorado State Univ., Fort Collins, CO (United States); Miller, Karen A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Flynn, Eric B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-10

    The objective of safeguards is the timely detection of the diversion of a significant quantity of nuclear materials, and safeguarding uranium enrichment plants is especially important in preventing the spread of nuclear weapons. The IAEA’s proposed Unattended Cylinder Verification Station (UCVS) for UF6 cylinder verification would combine the operator’s accountancy scale with a nondestructive assay system such as the Passive Neutron Enrichment Meter (PNEM) and cylinder identification and surveillance systems. In this project, we built a laboratory-scale UCVS and demonstrated its capabilities using mock UF6 cylinders. We developed a signal processing algorithm to automate the data collection and processing from four continuous, unattended sensors. The laboratory demonstration of the system showed that the software could successfully identify cylinders, snip sensor data at the appropriate points in time, determine the relevant characteristics of the cylinder contents, check for consistency among sensors, and output the cylinder data to a file. This paper describes the equipment, algorithm and software development, laboratory demonstration, and recommendations for a full-scale UCVS.

  4. Nuclear and thermal-hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-In; Kim, Young-Gyun; Kim, Sang-Ji; Kim, Young-Jin

    1999-05-01

    As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.

  5. Uranium Enrichment Determination Using a New Analysis Code for the U XKα Region: HyperGam-U

    Directory of Open Access Journals (Sweden)

    Junhyuck Kim

    2016-06-01

    Full Text Available HyperGam-U was recently developed to determine uranium enrichment based on γ- and X-ray spectroscopy analysis. The XKα region of the uranium spectrum contains 13 peaks for 235U and 238U and is used mainly for analysis. To describe the X-ray peaks, a Lorentzian broadened shape function was used, and methods were developed to reduce the number of fitting parameters for decomposing the strongly overlapping peaks using channel-energy, energy-width, and energy-efficiency calibration functions. For validation, eight certified reference material uranium samples covering uranium enrichments from 1% to 99% were measured using a high-resolution planar high-purity germanium detector and analyzed using the HyperGam-U code. When corrections for the attenuation and true coincidence summing were performed for the detection geometry in this experiment, the goodness of fit was improved by a few percent. The enrichment bias in this study did not exceed 2% compared with the certified values for all measured samples.

  6. Estimation of uranium and cobalt-60 distribution coefficients and uranium-235 enrichment at the Combustion Engineering Company site in Windsor, Connecticut

    International Nuclear Information System (INIS)

    Wang, Y.; Orlandini, K.A.; Yu, C.

    1996-05-01

    Site-specific distribution coefficients for uranium isotopes and cobalt-60 (Co-60) and the fraction of uranium-235 (U-235) enrichment by mass were estimated for environmental samples collected from the Combustion Engineering Company site in Windsor, CT. This site has been identified for remedial action under the US Department of Energy's (DOE) Formerly Utilized Sites Remedial Action Program. The authority of DOE at the Combustion Engineering site is limited to (1) Building 3; (2) other activities or areas associated exclusively with Building 3 (such as sewer lines); or (3) contamination that is exclusively highly enriched uranium. In this study, 16 samples were collected from the Combustion Engineering site, including 8 soil, 4 sediment, 3 water, and 1 water plus sludge sample. These samples were analyzed for isotopic uranium by alpha spectrometry and for Co-60 by gamma spectrometry. The site-specific distribution coefficient for each isotope was estimated as the ratio of extractable radionuclide activity in the solid phase to the activity in the contact solution following a 19-day equilibration. The uranium activity measurements indicate that uranium-234 (U-234) and uranium-238 (U-238) were in secular equilibrium in two soil samples and that soil and sediment samples collected from other sampling locations had higher U-234 activity than U-238 activity in both the solid and solution phases. The site-specific distribution coefficient (Kd) ranged from 82 to 44,600 mL/g for U-238 and from 102 to 65,900 mL/g for U-234. Calculation of U-235 enrichment by mass indicated that four soil samples had values greater than 0.20; these values were 0.37, 0.38, 0.46, and 0.68. Cobalt-60 activity was detected in only three sediment samples. The measured Co-60 activity in the solid phase ranged from 0.15 to 0.45 pCi/g and that in the water phase of all three samples combined was 4 pCi/L. The Kd value for Co-60 in the site brook sediment was calculated to be 70 mL/g

  7. Report of the Comptroller General of the United States. Comments on selected aspects of the Administration's proposal for government assistance to private uranium enrichment groups

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents GAO's views on matters raised by the Vice Chairman, Joint Committee on Atomic Energy, related to the private ownership of uranium enrichment and reemphasizes GAO's belief that the Government should add on to one of its existing enrichment plants to provide the next increment of enrichment capacity

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-30

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

  10. Development of rolled joint detachment system for low enriched uranium based advanced heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chatterjee, S., E-mail: subrata@barc.gov.in; Madhusoodanan, K.

    2014-07-01

    Highlights: • Life limiting mechanism of low enriched uranium based advanced heavy water reactor. • The coolant channels need to be designed as easily replaceable for long life of the reactor. • To meet the above design feature rolled joint detachment system has been developed. • The system works on the principle of shock heating and fast cooling. • The paper describes the design basis and observations from the experimental trials. - Abstract: Advanced heavy water reactor-low enriched uranium (AHWR-LEU) is designed for high discharge burnup and large scale commercial utilization of thorium with LEU as an external feed in a once through fuel cycle. It is a 300 MWe vertical pressure tube type, thorium based reactor, cooled by boiling light water and moderated by heavy water. The reactor incorporates a number of passive safety features and is associated with a fuel cycle having reduced environmental impact and offers more proliferation resistance compared to contemporary light water reactors (LWRs). Some distinguishing features which make AHWR-LEU unique from other conventional nuclear power reactors are core heat removal by natural circulation, negative void coefficient of reactivity under all event scenarios, easily replaceable pressure tubes, long design life, passive safety features, etc. The long design life of the reactor is possible to be met only when the pressure tube which has a limited life of 15–20 full power years (FPYs) due to a life limiting mechanism called irradiation enhanced creep and growth is easily replaceable. Pressure tube is joined at its ends to SS 403 end fittings through rolled joints. A system has been designed and developed for detachment of pressure tubes from the end fitting and its associated hardware without disturbing the surrounding components or coolant channels so that the end fitting can be reused. The system works on the principle of shock heating and fast cooling of the pressure tube in the rolled joint region and

  11. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  12. 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.

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

    International Nuclear Information System (INIS)

    Guimaraes, Leonam dos Santos; Degrange, Jean Pierre

    1998-01-01

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

  14. 77 FR 7128 - Low Enriched Uranium From France: Preliminary Results of Antidumping Duty Changed Circumstances...

    Science.gov (United States)

    2012-02-10

    ... remediation of the situation. In light of the disaster that struck Japan after entry of this merchandise into... nuclear fuel rods or assemblies. Natural uranium concentrates (U 3 O 8 ) with a U\\235\\ concentration of no greater than 0.711 percent and natural uranium concentrates converted into uranium hexafluoride with a U...

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

  17. Issues and recommendations related to replacement of CFC-114 at the uranium enrichment gaseous diffusion plant

    International Nuclear Information System (INIS)

    Anderson, B.L.; Banaghan, E.

    1993-01-01

    The operating uranium enrichment gaseous diffusion plants (GDPs) in Portsmouth, Ohio and Paducah, Kentucky, which are operated for the United States Department for Energy by Martin Marietta Energy Systems (MMES), currently use a chlorofluorocarbon (CFC-114) as the primary process stream coolant. Due to recent legislation embodied in the Clean Air Act, the production of this and other related chlorofluorocarbons (CFCS) are to be phased out with no production occurring after 1995. Since the plants lose approximately 500,000 pounds per year of this process stream coolant through various leaks, the GDPs are faced with the challenge of identifying a replacement coolant that will allow continued operation of the plants. MMES formed the CFC Task Team to identify and solve the various problems associated with identifying and implementing a replacement coolant. This report includes a review of the work performed by the CFC Task Team, and recommendations that were formulated based on this review and upon original work. The topics covered include; identifying a replacement coolant, coolant leak detection and repair efforts, coolant safety concerns, coolant level sensors, regulatory issues, and an analytical decision analysis

  18. Safeguards Guidance for Designers of Commercial Nuclear Facilities – International Safeguards Requirements for Uranium Enrichment Plants

    Energy Technology Data Exchange (ETDEWEB)

    Philip Casey Durst; Scott DeMuth; Brent McGinnis; Michael Whitaker; James Morgan

    2010-04-01

    For the past two years, the United States National Nuclear Security Administration, Office of International Regimes and Agreements (NA-243), has sponsored the Safeguards-by-Design Project, through which it is hoped new nuclear facilities will be designed and constructed worldwide more amenable to nuclear safeguards. In the course of this project it was recognized that commercial designer/builders of nuclear facilities are not always aware of, or understand, the relevant domestic and international safeguards requirements, especially the latter as implemented by the International Atomic Energy Agency (IAEA). To help commercial designer/builders better understand these requirements, a report was prepared by the Safeguards-by-Design Project Team that articulated and interpreted the international nuclear safeguards requirements for the initial case of uranium enrichment plants. The following paper summarizes the subject report, the specific requirements, where they originate, and the implications for design and construction. It also briefly summarizes the established best design and operating practices that designer/builder/operators have implemented for currently meeting these requirements. In preparing the subject report, it is recognized that the best practices are continually evolving as the designer/builder/operators and IAEA consider even more effective and efficient means for meeting the safeguards requirements and objectives.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  20. Active neutron and gamma-ray imaging of highly enriched uranium for treaty verification.

    Science.gov (United States)

    Hamel, Michael C; Polack, J Kyle; Ruch, Marc L; Marcath, Matthew J; Clarke, Shaun D; Pozzi, Sara A

    2017-08-11

    The detection and characterization of highly enriched uranium (HEU) presents a large challenge in the non-proliferation field. HEU has a low neutron emission rate and most gamma rays are low energy and easily shielded. To address this challenge, an instrument known as the dual-particle imager (DPI) was used with a portable deuterium-tritium (DT) neutron generator to detect neutrons and gamma rays from induced fission in HEU. We evaluated system response using a 13.7-kg HEU sphere in several configurations with no moderation, high-density polyethylene (HDPE) moderation, and tungsten moderation. A hollow tungsten sphere was interrogated to evaluate the response to a possible hoax item. First, localization capabilities were demonstrated by reconstructing neutron and gamma-ray images. Once localized, additional properties such as fast neutron energy spectra and time-dependent neutron count rates were attributed to the items. For the interrogated configurations containing HEU, the reconstructed neutron spectra resembled Watt spectra, which gave confidence that the interrogated items were undergoing induced fission. The time-dependent neutron count rate was also compared for each configuration and shown to be dependent on the neutron multiplication of the item. This result showed that the DPI is a viable tool for localizing and confirming fissile mass and multiplication.

  1. Portsmouth Uranium Enrichment Complex environmental monitoring report for calendar year 1984

    International Nuclear Information System (INIS)

    1985-05-01

    At the Portsmouth Uranium Enrichment Complex all effluent streams are sampled regularly and analyzed to assess compliance with applicable environmental standards. Radioactivity is measured in air, water, fish, produce, soil, and sediments; and radiation doses to the public are calculated. The calculated public radiation doses from process effluents are within US Department of Energy (DOE) and US Environmental Protection Agency standards. An extensive amount of engineering effort is underway to modify existing airborne radioactivity emissions sources to further reduce emissions in compliance with DOE policy. Non-radioactive effluents either presently comply with federal standards or will comply upon completion of planned projects. CY-1984 was the fourth full year under the current National Pollutant Discharge Elimination System (NPDES) permit for liquid effluents, and the second year the Ohio Environmental Protection Agency (EPA) assumed responsibility for administering the NPDES program. Compliance with the permit's discharge limits, with the exception of violations due to the startup of two new wastewater treatment facilities, did not present any significant problems. Overall compliance was 94.7%

  2. Portsmouth Uranium Enrichment Complex environmental monitoring report for calendar year 1985

    International Nuclear Information System (INIS)

    1986-05-01

    At the Portsmouth Uranium Enrichment Complex all effluent streams are sampled regularly and analyzed to assess compliance with applicable environmental standards. Radioactivity is measured in air, water, fish, produce, soil, sediments and vegetation; and radiation doses to the public are calculated. The calculated public radiation doses from process effluents are within US Department of Energy (DOE) and US Environmental protection Agency (USEPA) standards. An extensive amount of engineering effort is underway to modify existing airborne radioactivity emissions sources to further reduce emissions in compliance with DOE policy. Non-radioactive effluents either presently comply with federal standards or will comply upon completion of planned projects. CY-1985 was the fifth full year under the current National Pullutant Discharge Elimination System (NPDES) permit for liquid effluents, and the third year since the Ohio Environmental Protection Agency (OEPA) assumed responsibility for administering the NPDES program. Compliance with the permit's discharge limits, with the exception of violations due to the Coal Pile Treatment Facility and the Biodenitrification Plant for which projects to improve performance are currently underway, did not present any significant problems. Overall compliance was 95.5%. There was increased activity in the areas of groundwater monitoring, hazardous waste and PCB waste management to comply with Federal regulations

  3. Stationary and protable instruments for assay of HEU [highly enriched uranium] solids holdup

    International Nuclear Information System (INIS)

    Russo, P.A.; Sprinkle, J.K. Jr.; Stephens, M.M.; Brumfield, T.L.; Gunn, C.S.; Watson, D.R.

    1987-01-01

    Two NaI(Tl)-based instruments, one stationary and one portable, designed for automated assay of highly enriched uranium (HEU) solids holdup, are being evaluated at the scrap recovery facility of the Oak Ridge Y-12 Plant. The stationary instrument, a continuous monitor of HEU within the filters of the chip burner exhaust system, measures the HEU deposits that accumulate erratically and rapidly during chip burner operation. The portable system was built to assay HEU in over 100 m of elevated piping used to transfer UO 3 , UO 2 , and UF 4 powder to, from, and between the fluid bed conversion furnances and the powder storage hoods. Both instruments use two detector heads. Both provide immediate automatic readout of accumulated HEU mass. The 186-keV 235 U gamma ray is the assay signature, and the 60-keV gamma ray from an 241 Am source attached to each detector is used to normalize the 186-keV rate. The measurement geometries were selected for compatibility with simple calibration models. The assay calibrations were calculated from these models and were verified and normalized with measurements of HEU standards built to match geometries of uniform accumulations on the surfaces of the process equipment. This instrumentation effort demonstrates that simple calibration models can often be applied to unique measurement geometries, minimizing the otherwise unreasonable requirements for calibration standards and allowing extension of the measurements to other process locations

  4. 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

  5. Activities in Argentina related to the use of slightly enriched uranium in heavy water reactor NPPs

    International Nuclear Information System (INIS)

    Corcuera, R.

    1999-01-01

    An overview of activities related to the use of Slightly Enriched Uranium (SEU) fuel in HWR type NPPs, currently under execution in Argentina, is presented. The activities here described cover certain R and D lines as well as the main aspects of the Project 'Transition from full Natural-U to full SEU core in Atucha-I NPP'. Concerning the R and D lines, a summary is given on investigations related to reduction of void-coefficient using SEU fuel assemblies, annular pellet SEU fuel for bundle power flattening, etc. The main aspects of the above mentioned Project are outlined. At present, Atucha-I core is approaching a 40% of core load with SEU fuel, while the target of full SEU fuel core should be reached in 2-3 years. The expected exit burnup for such a core, namely 11000 MWD/tnU, is already currently obtained for SEU fuel in the present mixed core, while an increase in exit burnup of Natural-U fuel has also been obtained in very good agreement with reactor physics calculations. The comprehensive safety analysis carried out for each phase of the Project showed very moderated changes in plant behaviour under the set of postulated accidents and abnormal transients. A recent development, namely the CARA Project, aimed at unifying manufacturing of fuel assemblies for both operating NPPs in Argentina is presented in an accompanying paper. (author)

  6. 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)

  7. Recommendations to the NRC on acceptable standard format and content for the Fundamental Nuclear Material Control (FNMC) Plan required for low-enriched uranium enrichment facilities

    International Nuclear Information System (INIS)

    Moran, B.W.; Belew, W.L.; Hammond, G.A.; Brenner, L.M.

    1991-11-01

    A new section, 10 CFR 74.33, has been added to the material control and accounting (MC ampersand A) requirements of 10 CFR Part 74. This new section pertains to US Nuclear Regulatory Commission (NRC)-licensed uranium enrichment facilities that are authorized to produce and to possess more than one effective kilogram of special nuclear material (SNM) of low strategic significance. The new section is patterned after 10 CFR 74.31, which pertains to NRC licensees (other than production or utilization facilities licensed pursuant to 10 CFR Part 50 and 70 and waste disposal facilities) that are authorized to possess and use more than one effective kilogram of unencapsulated SNM of low strategic significance. Because enrichment facilities have the potential capability of producing SNM of moderate strategic significance and also strategic SNM, certain performance objectives and MC ampersand A system capabilities are required in 10 CFR 74.33 that are not contained in 10 CFR 74.31. This document recommends to the NRC information that the licensee or applicant should provide in the fundamental nuclear material control (FNMC) plan. This document also describes methods that should be acceptable for compliance with the general performance objectives. While this document is intended to cover various uranium enrichment technologies, the primary focus at this time is gas centrifuge and gaseous diffusion

  8. NUCLEAR ISOTOPIC DILUTION OF HIGHLY ENRICHED URANIUM BY DRY BLENDING VIA THE RM-2 MILL TECHNOLOGY

    International Nuclear Information System (INIS)

    Rajamani, Raj K.; Latchireddi, Sanjeeva; Devrani, Vikas; Sethi, Harappan; Henry, Roger; Chipman, Nate

    2003-01-01

    DOE has initiated numerous activities to focus on identifying material management strategies to disposition various excess fissile materials. In particular the INEEL has stored 1,700 Kg of offspec HEU at INTEC in CPP-651 vault facility. Currently, the proposed strategies for dispositioning are (a) aqueous dissolution and down blending to LEU via facilities at SRS followed by shipment of the liquid LEU to NFS for fabrication into LWR fuel for the TVA reactors and (b) dilution of the HEU to 0.9% for discard as a waste stream that would no longer have a criticality or proliferation risk without being processed through some type of enrichment system. Dispositioning this inventory as a waste stream via aqueous processing at SRS has been determined to be too costly. Thus, dry blending is the only proposed disposal process for the uranium oxide materials in the CPP-651 vault. Isotopic dilution of HEU to typically less than 20% by dry blending is the key to solving the dispositioning issue (i.e., proliferation) posed by HEU stored at INEEL. RM-2 mill is a technology developed and successfully tested for producing ultra-fine particles by dry grinding. Grinding action in RM-2 mill produces a two million-fold increase in the number of particles being blended in a centrifugal field. In a previous study, the concept of achieving complete and adequate blending and mixing (i.e., no methods were identified to easily separate and concentrate one titanium compound from the other) in remarkably short processing times was successfully tested with surrogate materials (titanium dioxide and titanium mono-oxide) with different particle sizes, hardness and densities. In the current project, the RM-2 milling technology was thoroughly tested with mixtures of natural uranium oxide (NU) and depleted uranium oxide (DU) stock to prove its performance. The effects of mill operating and design variables on the blending of NU/DU oxides were evaluated. First, NU and DU both made of the same oxide

  9. NUCLEAR ISOTOPIC DILUTION OF HIGHLY ENRICHED URANIUM BY DRY BLENDING VIA THE RM-2 MILL TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Raj K. Rajamani; Sanjeeva Latchireddi; Vikas Devrani; Harappan Sethi; Roger Henry; Nate Chipman

    2003-08-01

    DOE has initiated numerous activities to focus on identifying material management strategies to disposition various excess fissile materials. In particular the INEEL has stored 1,700 Kg of offspec HEU at INTEC in CPP-651 vault facility. Currently, the proposed strategies for dispositioning are (a) aqueous dissolution and down blending to LEU via facilities at SRS followed by shipment of the liquid LEU to NFS for fabrication into LWR fuel for the TVA reactors and (b) dilution of the HEU to 0.9% for discard as a waste stream that would no longer have a criticality or proliferation risk without being processed through some type of enrichment system. Dispositioning this inventory as a waste stream via aqueous processing at SRS has been determined to be too costly. Thus, dry blending is the only proposed disposal process for the uranium oxide materials in the CPP-651 vault. Isotopic dilution of HEU to typically less than 20% by dry blending is the key to solving the dispositioning issue (i.e., proliferation) posed by HEU stored at INEEL. RM-2 mill is a technology developed and successfully tested for producing ultra-fine particles by dry grinding. Grinding action in RM-2 mill produces a two million-fold increase in the number of particles being blended in a centrifugal field. In a previous study, the concept of achieving complete and adequate blending and mixing (i.e., no methods were identified to easily separate and concentrate one titanium compound from the other) in remarkably short processing times was successfully tested with surrogate materials (titanium dioxide and titanium mono-oxide) with different particle sizes, hardness and densities. In the current project, the RM-2 milling technology was thoroughly tested with mixtures of natural uranium oxide (NU) and depleted uranium oxide (DU) stock to prove its performance. The effects of mill operating and design variables on the blending of NU/DU oxides were evaluated. First, NU and DU both made of the same oxide

  10. The elimination of chlorinated, chlorofluorocarbon, and other RCRA hazardous solvents from the Y-12 Plant's enriched uranium operations

    International Nuclear Information System (INIS)

    Johnson, D.H.; Patton, R.L.; Thompson, L.M.

    1990-01-01

    A major driving force in waste minimization within the plant is the reduction of mixed radioactive wastes associated with operations on highly enriched uranium. High enriched uranium has a high concentration of the uranium-235 isotope (up to 97.5% enrichment) and is radioactive, giving off alpha and low level gamma radiation. The material is fissionable with as little as two pounds dissolved in water being capable of producing a spontaneous chain reaction. For these reasons the material is processed in small batches or small geometries. Additionally, the material is completely recycled because of its strategic and monetary value. Since the early eighties, the plant has had an active waste minimization program which has concentrated on substitution of less hazardous solvents wherever possible. The following paper summarizes efforts in two areas - development of a water-based machining coolant to replace perchloroethylene and substitution of an aliphatic solvent to replace solvents producing hazardous wastes as defined by the Resource, Conservation, and Recovery Act (RCRA)

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

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

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

  12. 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

  13. 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

  14. Interim report of working group on development and examination of new material, high performance centrifuge technology, Advisory Committee on Nuclear Uranium Enrichment

    International Nuclear Information System (INIS)

    1988-01-01

    The industrialization of uranium enrichment in Japan has been advanced by the Power Reactor and Nuclear Fuel Development Corp. by centrifugal separation technology. In April, 1988, the partial operation of the uranium enrichment prototype plant by centrifugal separation process (200 t SWU/year) was begun in Ningyo Pass, Okayama Prefecture, and its full operation is expected in January, 1989. Based on this achievement, Japan Nuclear Fuel Industry Co., Ltd. advances the construction of a commercial uranium enrichment plant in Rokkasho Village, Aomori Prefecture, aiming at the start of operation around 1991. On the other hand, the environment surrounding the uranium enrichment business in Japan is extremely severe at present, and due to the excessive supply capacity of world uranium enrichment service and the recent rapid appreciation of yen, the further improvement of the economical efficiency of Japanese uranium enrichment business is demanded. The working group held four meetings since May, 1988, and evaluated the present status of the research and development of new material, high performance centrifuges, and investigated and discussed the method of advancing the research and development hereafter. The results are reported. (Kako, I.)

  15. Depleted and enriched uranium exposure quantified in former factory workers and local residents of NL Industries, Colonie, NY USA

    International Nuclear Information System (INIS)

    Arnason, John G.; Pellegri, Christine N.; Moore, June L.; Lewis-Michl, Elizabeth L.; Parsons, Patrick J.

    2016-01-01

    Background: Between 1958 and 1982, NL Industries manufactured components of enriched (EU) and depleted uranium (DU) at a factory in Colonie NY, USA. More than 5 metric tons of DU was deposited as microscopic DU oxide particles on the plant site and surrounding residential community. A prior study involving a small number of individuals (n=23) indicated some residents were exposed to DU and former workers to both DU and EU, most probably through inhalation of aerosol particles. Objectives: Our aim was to measure total uranium [U] and the uranium isotope ratios: 234 U/ 238 U; 235 U/ 238 U; and 236 U/ 238 U, in the urine of a cohort of former workers and nearby residents of the NLI factory, to characterize individual exposure to natural uranium (NU), DU, and EU more than 3 decades after production ceased. Methods: We conducted a biomonitoring study in a larger cohort of 32 former workers and 99 residents, who may have been exposed during its period of operation, by measuring Total U, NU, DU, and EU in urine using Sector Field Inductively Coupled Plasma - Mass Spectrometry (SF-ICP-MS). Results: Among workers, 84% were exposed to DU, 9% to EU and DU, and 6% to natural uranium (NU) only. For those exposed to DU, urinary isotopic and [U] compositions result from binary mixing of NU and the DU plant feedstock. Among residents, 8% show evidence of DU exposure, whereas none shows evidence of EU exposure. For residents, the [U] geometric mean is significantly below the value reported for NHANES. There is no significant difference in [U] between exposed and unexposed residents, suggesting that [U] alone is not a reliable indicator of exposure to DU in this group. Conclusions: Ninety four percent of workers tested showed evidence of exposure to DU, EU or both, and were still excreting DU and EU decades after leaving the workforce. The study demonstrates the advantage of measuring multiple isotopic ratios (e.g., 236 U/ 238 U and 235 U/ 238 U) over a single ratio ( 235 U/ 238 U

  16. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  17. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  18. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  19. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  20. Study on developing brain damage of neonatal rats induced by enriched uranium

    International Nuclear Information System (INIS)

    Gu Guixiong; Zhu Shoupeng; Yang Shuqin

    2000-01-01

    Objective: The injurious effects of enriched uranium 235 U on developing brain of neonatal Wistar pure bred rats were studied. Methods: The model of irradiation induced brain damage in vivo was settled. The effects of cerebrum exposure by 235 U on somatic growth and neuro-behavior development of neonatal rats were examined by thirteen index determination of multiple parameters. The dynamic retention of autoradiographic tracks of 235 U in cells of developing brain was observed. The changes of NSE, IL-1β, SOD, and ET in cerebral cortex, hippocampus, diencephalon, cerebellum after expose to 235 U were examined with radioimmunoassay. Results: The somatic growth such as increase of body weight and brain weight was lower significantly. The retardation of development was found such as eye opening, sensuous function as auditory startle, movement and coordination function and activity as swimming, physiological reflexes as negative geotaxis, surface righting, grasping reflex suspension and the tendency behavior. The data showed delayed growth and abnormal neuro-behavior. The micro-autoradiographic tracing showed that the tracks of 235 U were mainly accumulated in the nucleus of developing brain. At the same time only few tracks appeared in the cytoplasm and interval between cells. Experimental study showed that when the dose of 235 U irradiation was increased, the level of NSE was decreased and the IL-1β was increased. However, the results indicated that SOD and ET can be elevated by the low dose irradiation of 235 U, and can be inhibited by the high dose. Conclusion: The behavior of internal irradiation from 235 U on the developing brain damage of neonatal rats were of sensibility and compensation in nervous cells

  1. Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1996-01-01

    A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution's concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the 'Poisoned Tube Tank' because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service

  2. Immobilization of uranium by biomaterial stabilized FeS nanoparticles: Effects of stabilizer and enrichment mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Dadong [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031 (China); Ren, Xuemei, E-mail: renxm_nana@163.com [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031 (China); Wen, Jun; Hu, Sheng; Xiong, Jie; Jiang, Tao [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Wang, Xiaolin, E-mail: xlwang@caep.ac.cn [China Academy of Engineering Physics, Mianyang 621900 (China); Wang, Xiangke, E-mail: xkwang@ncepu.edu.cn [School of Environment and Chemical Engineering, North China Electric Power University, Beijing 102206 (China)

    2016-01-25

    Highlights: • FeS can be stabilized by sodium carboxymethyl cellulose and gelatin. • Gelatin–FeS show best performance in U(VI) removal. • Adsorption–reduction immobilization of U(VI) is the major mechanism. - Abstract: Iron sulfide (FeS) nanoparticles have been recognized as effective scavengers for multi-valent metal ions. However, the aggregation of FeS nanoparticles in aqueous solution greatly restricts their application in real work. Herein, different biomaterial-FeS nanoparticles were developed for the in-situ immobilization of uranium(VI) in radioactive waste management. TEM images suggested that sodium carboxymethyl cellulose (CMC) and gelatin can effectively suppress the aggregation of FeS nanoparticles in aqueous solutions. The resulting CMC–FeS and gelatin–FeS were stable in aqueous solutions and showed high adsorption capacity for U(VI). Specially, gelatin–FeS showed the best performance in U(VI) adsorption–reduction immobilization under experimental conditions. The maximum enrichment capacity of U(VI) on CMC–FeS and gelatin–FeS at pH 5.0 and 20 °C achieved to ∼430 and ∼556 mg/g, respectively. Additionally, gelatin–FeS and CMC–FeS nanoparticles presented excellent tolerance to environmental salinity. The immobilized U(VI) on the surfaces of CMC–FeS and gelatin–FeS remained stable more than one year. These findings highlight the possibility of using ggelatin–FeS for efficient immobilization of U(VI) from radioactive wastewater.

  3. Anion-exchange enrichment and spectrophotometric determination of uranium in sea-water

    International Nuclear Information System (INIS)

    Kuroda, Rokuro; Oguma, Koichi; Mukai, Noriko; Iwamoto, Masatoshi

    1987-01-01

    A method is proposed for the determination of uranium in sea-water. The uranium is strongly sorbed on a strongly basic anion-exchange resin (Cl - form) from acidified sea-water containing sodium azide (0.3M) and is easily eluted with 1M hydrochloric acid. Uranium in the effluent can be determined spectrophotometrically with Arsenazo III. The combined method allows easy and selective determination of uranium in sea-water without using a sophisticated adsorbent. The overall recovery and precision are satisfactory at the 3 μg/1. level. (author)

  4. Monte Carlo modeling and analyses of YALINA booster subcritical assembly, Part III : low enriched uranium conversion analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y. (Nuclear Engineering Division)

    2011-05-12

    This study investigates the performance of the YALINA Booster subcritical assembly, located in Belarus, during operation with high (90%), medium (36%), and low (21%) enriched uranium fuels in the assembly's fast zone. The YALINA Booster is a zero-power, subcritical assembly driven by a conventional neutron generator. It was constructed for the purpose of investigating the static and dynamic neutronics properties of accelerator driven subcritical systems, and to serve as a fast neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinides. The first part of this study analyzes the assembly's performance with several fuel types. The MCNPX and MONK Monte Carlo codes were used to determine effective and source neutron multiplication factors, effective delayed neutron fraction, prompt neutron lifetime, neutron flux profiles and spectra, and neutron reaction rates produced from the use of three neutron sources: californium, deuterium-deuterium, and deuterium-tritium. In the latter two cases, the external neutron source operates in pulsed mode. The results discussed in the first part of this report show that the use of low enriched fuel in the fast zone of the assembly diminishes neutron multiplication. Therefore, the discussion in the second part of the report focuses on finding alternative fuel loading configurations that enhance neutron multiplication while using low enriched uranium fuel. It was found that arranging the interface absorber between the fast and the thermal zones in a circular rather than a square array is an effective method of operating the YALINA Booster subcritical assembly without downgrading neutron multiplication relative to the original value obtained with the use of the high enriched uranium fuels in the fast zone.

  5. Homogeneity study of enriched uranium dioxide to be used in intercomparison programs in the brazilian nuclear facilities

    International Nuclear Information System (INIS)

    Cristiano, Barbara F.G.; Lopes, Ricardo Tadeu

    2013-01-01

    The Brazilian nuclear facilities must dispose of measurement systems that allow the determination of their nuclear material inventories. The main goal of this work is the characterization of enriched uranium dioxide (UO 2 ) pellets to use as test sample to evaluate and verify these systems through the participation of the involved laboratories in interlaboratorial comparison programs. The programs are formed by a network of specialized laboratories in determining materials of the nuclear fuel cycle. Therefore, before being sent to laboratories, the materials must have their homogeneity confirmed. The homogeneity study of the UO 2 pellets has being carried out in the Safeguards Laboratory (LASAL) of Brazilian Nuclear Energy Commission (CNEN) through Destructives Analysis technique. The technique used was Davies and Gray/NBL titrimetric method. For this purpose, 25 pellets of enriched uranium dioxide from two different batches were randomly chosen. The model for homogeneity study between-unit variation (between-bottle homogeneity study) and minimum sample intake (within-bottle homogeneity study) adopted is a one-way ANOVA. No statistical significant differences were observed in the results of total uranium concentration for both batches. Thus, the UO 2 pellets are considered homogeneous and can be used in a Brazilian measurement systems evaluation program.(author)

  6. Development of ISA procedure for uranium fuel fabrication and enrichment facilities: overview of ISA procedure and its application

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki

    2013-01-01

    Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)

  7. Mineralization enrichment of uranyl mineral in mid-low temperature hydrothermal of Shuanghuajiang uranium deposit

    International Nuclear Information System (INIS)

    Fang Shiyi; Liang Yongdong; Tao Zhijun; Chen Weifeng

    2009-01-01

    Based on field examination and mineral study by X-ray powder diffraction, electronic probe and back scatter electronic image, uranium mineral in Shuanghuajiang uranium deposit are analyzed and the mineral are divided into two classes: uranyl phosphate and uranyl silicate. After studying the forming condition of these uranylminerals, the deposit fall into low temperature hydrothermal genesis. The metallogenic model is hence established. (authors)

  8. Extrapolated experimental critical parameters of unreflected and steel-reflected massive enriched uranium metal spherical and hemispherical assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rothe, R.E.

    1997-12-01

    Sixty-nine critical configurations of up to 186 kg of uranium are reported from very early experiments (1960s) performed at the Rocky Flats Critical Mass Laboratory near Denver, Colorado. Enriched (93%) uranium metal spherical and hemispherical configurations were studied. All were thick-walled shells except for two solid hemispheres. Experiments were essentially unreflected; or they included central and/or external regions of mild steel. No liquids were involved. Critical parameters are derived from extrapolations beyond subcritical data. Extrapolations, rather than more precise interpolations between slightly supercritical and slightly subcritical configurations, were necessary because experiments involved manually assembled configurations. Many extrapolations were quite long; but the general lack of curvature in the subcritical region lends credibility to their validity. In addition to delayed critical parameters, a procedure is offered which might permit the determination of prompt critical parameters as well for the same cases. This conjectured procedure is not based on any strong physical arguments.

  9. Converting targets and processes for fission-product molybdenum-99 from high- to low-enriched uranium

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Snelgrove, J.L.; Aase, S.

    1999-01-01

    Most of the world's supply of 99 Mo is produced by the fissioning of 235 U in high-enriched uranium targets (HEU, generally 93% 235 U). To reduce nuclear-proliferation concerns, the U.S. Reduced Enrichment for Research and Test Reactor Program is working to convert the current HEU targets to low-enriched uranium (LEU, 235 U). Switching to LEU targets also requires modifying the separation processes. Current HEU processes can be classified into two main groups based on whether the irradiated target is dissolved in acid or base. Our program has been working on both fronts, with development of targets for acid-side processes being the furthest along. However, using an LEU metal foil target may allow the facile replacement of HEU for both acid and basic dissolution processes. Demonstration of the irradiation and 99 Mo separation processes for the LEU metal-foil targets is being done in cooperation with researchers at the Indonesian PUSPIPTEK facility. We are also developing LEU UO 2 /Al dispersion plates as substitutes for HEU UA1 x /A1 dispersion plates for base-side processes. Results show that conversion to LEU is technically feasible; working with producers is essential to lowering any economic penalty associated with conversion. (author)

  10. H.R. 788: This Act may be cited as the Uranium Enrichment Reorganization Act, introduced in the House of Representatives, One Hundred Second Congress, First Session, February 4, 1991

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    This bill would maintain a competitive, financially strong, and secure uranium enrichment capability in the US by reorganizing the uranium enrichment enterprise. The bill amends the Atomic Energy Act of 1954 by establishing the United States Uranium Enrichment Corporation. This bill describes general provisions; the establishment of the corporation; powers and duties of the corporation; organization, finance, and management; licensing, taxation, and miscellaneous provisions; decontamination, decommissioning, and remedial action

  11. The fluorimetry for control of internal contamination of exposed workers to natural and enriched uranium

    International Nuclear Information System (INIS)

    Gaburo, J.C.; Todo, A.S.; Sordi, G.M.A.A.

    2000-01-01

    This study is a part of bioassay program revision applied to the uranium processing plants at IPEN-CNEN/SP. The workers of these facilities handle both natural uranium and uranium compounds with different isotopic composition which could reach up to 20% in 235 U. The most commonly employed techniques for the determination of uranium in urine at IPEN are fluorimetry and alpha spectrometry with detection limit of 1.0 mgL-1. and 1,0 mBqL-1 , respectively. Based in advantages and disadvantages of each technique it is very important to identify the workers groups that should be submitted for these analysis. In this report a limiting value of uranium concentration in urine, mgL-1, obtained by fluorimetry is proposed. All the results greater than these limiting value indicate the necessity to carry out a additional measurement by alpha spectroscopy. The uranium mass that result in a pre-determined limit committed effective dose is function of isotopic composition. Consequently, the predicted value of the measured of urinary excretion is function of isotopic composition also and depends of absorption characteristics when inhaled and of the monitoring interval considered. In this report the uranium concentration values for reference levels and limits doses are determined. Based on these results the procedures to use the fluorimetry or both fluorimetry and alpha-spectrometry were adopted. (author)

  12. Compton DIV: Using a Compton-Based Gamma-Ray Imager for Design Information Verification of Uranium Enrichment Plants

    International Nuclear Information System (INIS)

    Burks, M.; Verbeke, J.; Dougan, A.; Wang, T.; Decman, D.

    2009-01-01

    A feasibility study has been performed to determine the potential usefulness of Compton imaging as a tool for design information verification (DIV) of uranium enrichment plants. Compton imaging is a method of gamma-ray imaging capable of imaging with a 360-degree field of view over a broad range of energies. These systems can image a room (with a time span on the order of one hour) and return a picture of the distribution and composition of radioactive material in that room. The effectiveness of Compton imaging depends on the sensitivity and resolution of the instrument as well the strength and energy of the radioactive material to be imaged. This study combined measurements and simulations to examine the specific issue of UF 6 gas flow in pipes, at various enrichment levels, as well as hold-up resulting from the accumulation of enriched material in those pipes. It was found that current generation imagers could image pipes carrying UF 6 in less than one hour at moderate to high enrichment. Pipes with low enriched gas would require more time. It was also found that hold-up was more amenable to this technique and could be imaged in gram quantities in a fraction of an hour. another questions arises regarding the ability to separately image two pipes spaced closely together. This depends on the capabilities of the instrument in question. These results are described in detail. In addition, suggestions are given as to how to develop Compton imaging as a tool for DIV

  13. Benchmark critical experiments on low-enriched uranium oxide systems with H/U = 0.77

    International Nuclear Information System (INIS)

    Tuck, G.; Oh, I.

    1979-08-01

    Ten benchmark experiments were performed at the Critical Mass Laboratory at Rockwell International's Rocky Flats Plant, Golden, Colorado, for the US Nuclear Regulatory Commission. They provide accurate criticality data for low-enriched damp uranium oxide (U 3 O 8 ) systems. The core studied consisted of 152 mm cubical aluminum cans containing an average of 15,129 g of low-enriched (4.46% 235 U) uranium oxide compacted to a density of 4.68 g/cm 3 and with an H/U atomic ratio of 0.77. One hundred twenty five (125) of these cans were arranged in an approx. 770 mm cubical array. Since the oxide alone cannot be made critical in an array of this size, an enriched (approx. 93% 235 U) metal or solution driver was used to achieve criticality. Measurements are reported for systems having the least practical reflection and for systems reflected by approx. 254-mm-thick concrete or plastic. Under the three reflection conditions, the mass of the uranium metal driver ranged from 29.87 kg to 33.54 kg for an oxide core of 1864.6 kg. For an oxide core of 1824.9 kg, the weight of the high concentration (351.2 kg U/m 3 ) solution driver varied from 14.07 kg to 16.14 kg, and the weight of the low concentration (86.4 kg U/m 3 ) solution driver from 12.4 kg to 14.0 kg

  14. Spectrographic determination of impurities in enriched uranium solutions; Determinacion espectrografica de impurezas en soluciones de uranio enriquecido

    Energy Technology Data Exchange (ETDEWEB)

    Capdevila, C.; Roca, M.

    1980-07-01

    A spectrographic procedure for the determination of trace amounts of Al, B, Ba, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, K, L i , Hg, Mn, Mo, Na, Nb, Ni, P, Pb, Ru, Sb, Sn, Sr, Ti, V, Zn, and Zr in enriched uranyl nitrate solutions from the reprocessing of spent nuclear fuels is described. After removal of uranium by either TBP or TNOA solvent extraction, the aqueous phase Is analysed by the graphite spark technique. TBP is adequate for all impurities, excepting boron and phosphorus; both of these elements can sat is factory be determined by using TNOA after the addition of mannitol to avoid boron losses. (Author) 4 refs.

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

    International Nuclear Information System (INIS)

    Arbie, B.; Soentono, S.

    1987-01-01

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

  16. A preliminary study on financial risk management of the uranium enrichment company, given China's fast development of nuclear power industry

    International Nuclear Information System (INIS)

    Sun Wumei

    2010-01-01

    In market economy, control and prevention of financial risk is a factual and urgent issue of financial management. As an independent legal identity, the enterprise strives to achieve business operations and developments while bearing risks. In an era of fast growth of China's nuclear power industry, uncertainties facing uranium enrichment companies are inevitable. Without being aware of financial risks, the enterprise could be driven to hopeless situation when catastrophe comes. Recognizing financial risk factors, knowing the cause and effect of risk events, building a mechanism to prevent and control financial risks, are the right approaches for building a robust enterprise. (author)

  17. Chemical toxicity and radiological health detriment associated with the inhalation of various enrichments of uranium.

    Science.gov (United States)

    Bryant, P A

    2014-03-01

    The occupational risks associated with the chemical toxicity of uranium can be overlooked during the processing, handling and storage of the material, as the radioactivity of the material is often used alone to assess the health consequences of exposure to uranium compounds. This note provides a summary of the current United Kingdom occupational standards for uranium based on radiation dose and/or chemical toxicity with a particular focus on intake via inhalation. A simple model is subsequently presented to allow a comparison to be drawn between the occupational exposure standard for chemical toxicity and radiological dose limit. Using these data a set of suggested limits on occupational exposure to airborne uranium is proposed that indicate where the legal annual radiological dose limit for workers or the Health and Safety Executive occupational exposure standard for chemical toxicity are at risk of being breached.

  18. Chemical toxicity and radiological health detriment associated with the inhalation of various enrichments of uranium

    International Nuclear Information System (INIS)

    3C Limited, Queen Square House, 18-21 Queen Square, Bristol BS1 4NH (United Kingdom))" data-affiliation=" (SR3C Limited, Queen Square House, 18-21 Queen Square, Bristol BS1 4NH (United Kingdom))" >Bryant, P A

    2014-01-01

    The occupational risks associated with the chemical toxicity of uranium can be overlooked during the processing, handling and storage of the material, as the radioactivity of the material is often used alone to assess the health consequences of exposure to uranium compounds. This note provides a summary of the current United Kingdom occupational standards for uranium based on radiation dose and/or chemical toxicity with a particular focus on intake via inhalation. A simple model is subsequently presented to allow a comparison to be drawn between the occupational exposure standard for chemical toxicity and radiological dose limit. Using these data a set of suggested limits on occupational exposure to airborne uranium is proposed that indicate where the legal annual radiological dose limit for workers or the Health and Safety Executive occupational exposure standard for chemical toxicity are at risk of being breached. (note)

  19. 78 FR 17942 - Request To Amend a License To Export High-Enriched Uranium

    Science.gov (United States)

    2013-03-25

    ...%)) kilograms Czech Republic to medical isotope March 11, 2013 uranium) the list of production at the XSNM3622... Belgium for ultimate use as fuel or targets for medical isotope production Dated this 14th day of March...

  20. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

  1. Uranium

    International Nuclear Information System (INIS)

    Batley, G.C.; McKay, A.D.

    1986-01-01

    Production of uranium oxide in Australia for 1983 was 3786 t(3211 t U). Uranium exports for 1983 were 3273 t U 3 O 8 at an average f.o.b. value of $41.02/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1982-83 fiscal year was $36.5 million, 35% less than in 1981-82. In November 1983, the Government decided that uranium mining would be allowed only at the existing Ranger and Nabarlek mines and at the proposed Olympic Dam mine. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U as at December 1983, totalled 474 000 t U. Australia's total now represents 30% of the Western world's low-cost RAR. In addition Australia has 235 000 t U in the low-cost Estimated Additional Resources Category 1, which represents 31% of the Western world's resources in this category

  2. Application of the HGSYSTEM/UF6 model to simulate atmospheric dispersion of UF6 releases from uranium enrichment plants

    International Nuclear Information System (INIS)

    Goode, W.D. Jr.; Bloom, S.G.; Keith, K.D. Jr.

    1995-01-01

    Uranium hexafluoride is a dense, reactive gas used in Gaseous Diffusion Plants (GDPs) to make uranium enriched in the 235 U isotope. Large quantities of UF 6 exist at the GDPs in the form of in-process gas and as a solid in storage cylinders; smaller amounts exist as hot liquid during transfer operations. If liquid UF 6 is released to the environment, it immediately flashes to a solid and a dense gas that reacts rapidly with water vapor in the air to form solid particles of uranyl fluoride and hydrogen fluoride gas. Preliminary analyses were done on various accidental release scenarios to determine which scenarios must be considered in the safety analyses for the GDPS. These scenarios included gas releases due to failure of process equipment and liquid/gas releases resulting from a breach of transfer piping from a cylinder. A major goal of the calculations was to estimate the response time for mitigating actions in order to limit potential off-site consequences of these postulated releases. The HGSYSTEM/UF 6 code was used to assess the consequences of these release scenarios. Inputs were developed from release calculations which included two-phase, choked flow followed by expansion to atmospheric pressure. Adjustments were made to account for variable release rates and multiple release points. Superpositioning of outputs and adjustments for exposure time were required to evaluate consequences based on health effects due to exposures to uranium and HF at a specific location

  3. Managing proliferation risks from civilian and weapon-grade plutonium and enriched uranium: A comprehensive cut-off convention

    International Nuclear Information System (INIS)

    Liebert, W.

    1997-01-01

    The problem of weapon-grade fissile materials is closely related to the aim of achieving a nuclear-weapon-free world. Huge amounts of highly enriched uranium have been produced for nuclear weapons. More than 1000 tonnes of plutonium emerged as a by-product of civilian nuclear industry. Separated from spent fuel it is readily usable for nuclear weapons. The worldwide civilian tritium inventory may reach the same size as military stocks about the year 2010. This poses an increasing danger of horizontal nuclear proliferation. Production, stockpiling, trade, processing and uses of weapon-grade materials like Highly enriched uranium, plutonium and tritium promote its geographical spread, enlarge the group of people with the relate know-how and create the danger of diversion of material and the proliferation of knowledge for the purpose of weapons production. Therefore, a fundamental turn away from using weapon-grade materials in scientific and economic applications of nuclear energy is desirable in all countries. Priority should be given to using nuclear fuel cycles which are as proliferation resistant as possible. Without this, the continuation of civil nuclear programs seems to be irresponsible and unjustifiable. The role of the IAEA in export control safeguards related to the above problems is indispensable

  4. The NNSA global threat reduction initiative's efforts to minimize the use of highly enriched uranium for medical isotope production

    International Nuclear Information System (INIS)

    Staples, Parrish

    2010-01-01

    The mission of the National Nuclear Security Administration's (NNSA) Office of Global Threat Reduction (GTRI) is to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. GTRI is a key organization for supporting domestic and global efforts to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications. GTRI implements the following activities in order to achieve its threat reduction and HEU minimization objectives: Converting domestic and international civilian research reactors and isotope production facilities from the use of HEU to low enriched uranium (LEU); Demonstrating the viability of medical isotope production technologies that do not use HEU; Removing or disposing excess nuclear and radiological materials from civilian sites worldwide; and Protecting high-priority nuclear and radiological materials worldwide from theft and sabotage. This paper provides a brief overview on the recent developments and priorities for GTRI program activities in 2010, with a particular focus on GTRI's efforts to demonstrate the viability of non-HEU based medical isotope production technologies. (author)

  5. The Supply of Medical Radioisotopes. Market impacts of converting to low-enriched uranium targets for medical isotope production

    International Nuclear Information System (INIS)

    Westmacott, Chad; Cameron, Ron

    2012-01-01

    The reliable supply of molybdenum-99 ( 99 Mo) and its decay product, technetium-99m ( 99m Tc), is a vital component of modern medical diagnostic practices. At present, most of the global production of 99 Mo is from highly enriched uranium (HEU) targets. However, all major 99 Mo-producing countries have recently agreed to convert to using low-enriched uranium (LEU) targets to advance important non-proliferation goals, a decision that will have implications for the global supply chain of 99 Mo/ 99m Tc and the long-term supply reliability of these medical isotopes. This study provides the findings and analysis from an extensive examination of the 99 Mo/ 99m Tc supply chain by the OECD/NEA High-level Group on the Security of Supply of Medical Radioisotopes (HLG-MR). It presents a comprehensive evaluation of the potential impacts of converting to the use of LEU targets for 99 Mo production on the global 99 Mo/ 99m Tc market in terms of costs and available production capacity, and the corresponding implications for long-term supply reliability. In this context, the study also briefly discusses the need for policy action by governments in their efforts to ensure a stable and secure long-term supply of 99 Mo/ 99m Tc

  6. Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

    International Nuclear Information System (INIS)

    Muhammad, Farhan

    2011-01-01

    Highlights: → The isotopic concentration in the fuel changes as soon as it starts its operation. → The neutronic properties of a reactor also change with fuel burnup. → The reactivity feedbacks at end-of-life of a material test reactor fuelled with low enriched uranium fuel are calculated. → Codes used include WIMS-D4 and CITATION. - Abstract: The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA's 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2-5%.

  7. Laser and gas centrifuge enrichment

    Science.gov (United States)

    Heinonen, Olli

    2014-05-01

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

  8. Uranium enrichment in South Africa: from the world-unique Z-plant to the use of high-technology lasers

    International Nuclear Information System (INIS)

    McDowell, M.W.

    1995-01-01

    A historical discussion of the technology used in South Africa for the enrichment of uranium, as well as other technological spin-offs for the country that followed from the construction of the Z-plant. The national energy strategy and objectives of the government during the Apartheid years resulted in the development of several large-scale energy projects. The pressure of sanctions forced the Z-plant to be rushed into operation at an uneconomical capacity of 250 000 SWU per annum. In 1994 this implied that enriched uranium was produced at a cost of $200 per SWU while the world market price was below $90. While the production of enriched uranium at the Z-plant ceased early in 1995, the expertise gained will not be lost entirely. As a result of the high energy and financial capital intensive current methods of producing enriched uranium, research started in the early 1970's into alternative production processes making use of lasers. South Africa has opted for the MLIS (molecular laser isotope separation) process, as a result of its vast experience gained from the Z-plant in the handling of the molecular input gas UF6 (uranium hexafluoride), and this has been under development since the early 1980's. During 1994 significant progress was made with MLIS, in particular with single-step enrichment from natural uranium to better than 4% uranium 235 on a macro scale. The Atomic Energy Corporation of South Africa's strategy is to licence the process internationally. 3 tabs., 3 figs

  9. Calibration Tools for Measurement of Highly Enriched Uranium in Oxide and Mixed Uranium-Plutonium Oxide with a Passive-Active Neutron Drum Shuffler

    International Nuclear Information System (INIS)

    Mount, M; O'Connell, W; Cochran, C; Rinard, P

    2003-01-01

    Lawrence Livermore National Laboratory (LLNL) has completed an extensive effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. Earlier papers described the PAN shuffler calibration over a range of item properties by standards measurements and an extensive series of detailed simulation calculations. With a single normalization factor, the simulations agree with the HEU oxide standards measurements to within ±1.2% at one standard deviation. Measurement errors on mixed U-Pu oxide samples are in the ±2% to ±10% range, or ±20 g for the smaller items. The purpose of this paper is to facilitate transfer of the LLNL procedure and calibration algorithms to external users who possess an identical, or equivalent, PAN shuffler. Steps include (1) measurement of HEU standards or working reference materials (WRMs); (2) MCNP simulation calculations for the standards or WRMs and a range of possible masses in the same containers; (3) a normalization of the calibration algorithms using the standard or WRM measurements to account for differences in the 252 Cf source strength, the delayed-neutron nuclear data, effects of the irradiation protocol, and detector efficiency; and (4) a verification of the simulation series trends against like LLNL results. Tools include EXCEL/Visual Basic programs which pre- and post-process the simulations, control the normalization, and embody the calibration algorithms

  10. Uranium

    International Nuclear Information System (INIS)

    Villarreal, E.

    1986-01-01

    After the increase in oil prices in 1973, several European countries increased their power programs. As a result some uranium mining companies from the FRG, Spain and France invested in exploration of radioactive minerals in Colombia hoping to find uranium resources needed to fuel European reactors. In the article a historic review of foreign investment in uranium in Colombia is made; some recommendations about joint-venture contracts used to regulate the work of the foreign companies are included. The four companies involved in exploration left the country in the early eighties, due to the difficulties in finding a large deposit and the difficult world situation of nuclear power

  11. Comments by the delegation of Japan on the American paper entitled 'institutional arrangements for uranium enrichment'

    International Nuclear Information System (INIS)

    1979-02-01

    It is argued that surplus global enrichment capacity through at least 1990, as suggested in CC/WG.2/11, may or may not exist, and to some extent may be necessary as an assurance of supply. CC/WG.2/11 also gives insufficient attention to the relation between national enrichment facilities and assurance of supply. The creation of multi-national enrichment enterprises may contribute to the spread of sensitive enrichment technology, a point which is explored more fully in CC/WG.2/9

  12. Los Alamos National Laboratory Support for Commercial U.S. Production of 99Mo without the Use of Highly Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-04

    There is currently a serious shortage of 99Mo, from which to generate the medically significant isotope 99mTc. Most of the world's supply comes from the fission of highly enriched uranium targets--this is a proliferation concern. This document focuses on the technology involved in two alternative methods: electron accelerator production of 99Mo from the 100Mo(γ,n)99Mo reaction and production of 99Mo as a fission product in a subcritical, DT accelerator-driven low enriched uranium salt solution.

  13. Communication dated 19 May 2011 received from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency regarding Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2011-01-01

    The Secretariat has received a letter dated 19 May 2011 from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency, attaching the Proposal for the Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants, as described in document GOV/2011/10. As requested by the Resident Representative, the letter and its attachment are circulated herewith for information of all Member States [es

  14. Communication dated 19 May 2011 received from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency regarding Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2011-01-01

    The Secretariat has received a letter dated 19 May 2011 from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency, attaching the Proposal for the Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants, as described in document GOV/2011/10. As requested by the Resident Representative, the letter and its attachment are circulated herewith for information of all Member States

  15. A multi-instrumental geochemical study of anomalous uranium enrichment in coal

    Czech Academy of Sciences Publication Activity Database

    Havelcová, Martina; Machovič, Vladimír; Mizera, Jiří; Sýkorová, Ivana; Borecká, Lenka; Kopecký, L.

    2014-01-01

    Roč. 137, November (2014), s. 52-63 ISSN 0265-931X R&D Projects: GA ČR(CZ) GA13-18482S Institutional support: RVO:67985891 ; RVO:61389005 Keywords : uranium * coal * organic matter * Sokolov Basin Subject RIV: DD - Geochemistry Impact factor: 2.483, year: 2014

  16. Determination of the isotopic composition of natural and slightly enriched uranium by alpha-spectrometry

    DEFF Research Database (Denmark)

    Kunzendorf, Helmar

    1968-01-01

    Determinations of the isotope contents of 238U, 235U and 234U in five uranium samples containing 0–5 at% 235U were carried out on the basis of a least-squares fit of the α-spectra from the samples, measured with a semiconductor detector, to the theoretically expected α-spectra. With a simple source...

  17. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  19. The text of the Agreement of 30 September 1986 concerning the Agency's assistance to Thailand for the transfer of enriched uranium from the United States for a research reactor

    International Nuclear Information System (INIS)

    1987-03-01

    The document reproduces the text of the Project and Supply Agreement concluded on 30 September 1986 between the Agency and the Government of the Kingdom of Thailand and the United States of America for the transfer of enriched uranium (approximately 49.0 kilograms of uranium enriched to approximately 19.90 per cent by weight in the isotope uranium-235) for a research reactor (TRIGA Mark III) in Thailand

  20. 78 FR 72123 - Request To Amend a License to Export High-Enriched Uranium

    Science.gov (United States)

    2013-12-02

    .... Material type Total quantity End use country U.S. Department of Energy, High-Enriched 13.5 kilograms To... U- production of molybdenum-99 XSNM3729/01, 11006053. 235). (Mo-99) medical isotopes in the...

  1. Computer-optimized γ-NDA geometries for uranium enrichment verification of gaseous UF6

    International Nuclear Information System (INIS)

    Wichers, V.A.; Aaldijk, J.K.; Betue, P.A.C. de; Harry, R.J.S.

    1993-05-01

    An improved collimator pair of novel design tailored for deposit independent enrichment verification of gaseous UF 6 at low pressures in cascade-to-header pipes of small diameters in centrifuge enrichment plants is presented. The designs are adapted for use in a dual-geometry arrangement for simultaneous measurements with both detection geometries. The average measurement time with the dual-geometry arrangement is approximately half an hour for deposit-to-gas activity ratios as high as 20. (orig.)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  3. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  4. Uranium*

    Science.gov (United States)

    Grenthe, Ingmar; Drożdżyński, Janusz; Fujino, Takeo; Buck, Edgar C.; Albrecht-Schmitt, Thomas E.; Wolf, Stephen F.

    Uranium compounds have been used as colorants since Roman times (Caley, 1948). Uranium was discovered as a chemical element in a pitchblende specimen by Martin Heinrich Klaproth, who published the results of his work in 1789. Pitchblende is an impure uranium oxide, consisting partly of the most reduced oxide uraninite (UO2) and partly of U3O8. Earlier mineralogists had considered this mineral to be a complex oxide of iron and tungsten or of iron and zinc, but Klaproth showed by dissolving it partially in strong acid that the solutions yielded precipitates that were different from those of known elements. Therefore he concluded that it contained a new element (Mellor, 1932); he named it after the planet Uranus, which had been discovered in 1781 by William Herschel, who named it after the ancient Greek deity of the Heavens.

  5. Feeding the nuclear fuel cycle with a long term view; AREVA's front-end business units, uranium mining, UF6 conversion and isotopic enrichment

    International Nuclear Information System (INIS)

    Capus, G.A.P.; Autegert, R.

    2005-01-01

    As a leading provider of technological solutions for nuclear power generation and electricity transmission, the AREVA group has the unique capability of offering a fully integrated fuel supply, when requested by its customers. At the core of the AREVA group, COGEMA Front End Division is an essential part of the overall fuel supply chain. Composed of three Business Units and gathering several subsidiaries and joint 'ventures, this division enjoys several leading positions as shown by its market shares and historical production records. Current Uranium market evolutions put the natural uranium supply under focus. The uranium conversion segment also recently revealed some concerning evolutions. And no doubt, the market pressure will soon be directed also at the enrichment segment. Looking towards the long term, AREVA strongly believes that a nuclear power renewal is needed, especially to help limiting green house effect gas release. Therefore, to address future supplies needed to fuel the existing fleet of nuclear power plants, but also new ones, the AREVA group is planning very significant investments to build new facilities in all the three front-end market segments. As far as uranium mining is concerned, these new mines will be based upon uranium reserves of outstanding quality. As for uranium conversion and enrichment, two large projects will be based on the most advanced technologies. This paper is aimed at recalling COGEMA Front End Division experience, the current status of its plants and operating entities and will provide a detailed overview of its major projects. (authors)

  6. A multi-instrumental geochemical study of anomalous uranium enrichment in coal.

    Science.gov (United States)

    Havelcová, Martina; Machovič, Vladimír; Mizera, Jiří; Sýkorová, Ivana; Borecká, Lenka; Kopecký, Lubomír

    2014-11-01

    Contents of uranium in coals from Odeř in the northernmost part of the Sokolov Basin, Czech Republic, in the vicinity of the well known St. Joachimsthal uranium ore deposits, reach extremely high values. In the present work, coal samples with contents of uranium ranging from 0.02 to 6 wt.% were studied. The study employing a whole complex of analytical techniques has been aimed at identification of changes in the structure of coal organic matter, which are associated with the high contents of uranium in coal. The study includes proximate and ultimate analyses, multielement analysis by instrumental neutron and photon activation analyses, micropetrographic analysis by optical microscopy, ESEM/EDX analysis of mineral matter, infrared and Raman spectroscopies, solvent extraction followed by gas chromatography with mass spectroscopy (GC/MS), and analytical pyrolysis (Py-GC/MS). The study has confirmed previously proposed explanation of uraniferous mineralization in sedimentary carboniferous substances by the mechanism of reduction and fixation of soluble U(VI) (uranyl, UO2(2+)) species (e.g., humic, carbonate/hydroxo/phosphate complexes) by sedimentary organic matter under diagenetic or hydrothermal conditions, and formation of insoluble U(IV) species as phosphate minerals and uraninite. The process is accompanied with alteration and destruction of the coal organic matter. The changes in the structure of coal organic matter involve dehydrogenation and oxidation mainly in the aliphatic, aromatic and hydroxyl structures, and an increase in aromaticity, content of ether bonds, and the degree of coalification. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  8. Design and construction of a Type B overpack container for the safe transportation of enriched uranium hexafluoride

    International Nuclear Information System (INIS)

    Gablin, K.A.

    1976-01-01

    The Paducah Tiger is an overpack designed for the international shipment of ten-ton cylinders of uranium hexafluoride in enriched form above the level of low specific acitivity. This container is designed as a Type B Package and has undergone all the tests and analyses required for a U.S. Department of Transportation Permit No. 6553. The Paducah Tiger is currently being used to ship fuel material in the USA on both truck and rail modes of transportation. In many ways, the design resembles the Super Tigersup(R), but incorporates features such as ISO corners, quick opening fasteners, and interior shock isolators that provide a system approach to the high volume of fuel shipment required in the last half of the 20th century. (author)

  9. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched (<20% {sup 235}U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  10. Department of Energy, highly enriched uranium ES ampersand H vulnerability assessment, Idaho National Engineering Laboratory site assessment team report

    International Nuclear Information System (INIS)

    1996-01-01

    In accordance with the February 22, 1996 directive issued by Secretary of Energy O'Leary on the Vulnerability Assessment of Highly Enriched Uranium (HEU) Storage, the Idaho National Engineering Laboratory conducted an assessment of the site's HEU holdings and any associated vulnerabilities. The assessment was conducted between April 25 and May 24, 1996. The scope of this assessment, as defined in the Assessment Plan, included all HEU, and any spent fuel not evaluated in the Spent Fuel Vulnerability Assessment. Addressed in this assessment were all of the holdings at the Idaho National Engineering Laboratory (INEL) except any located at Argonne National Laboratory-West (ANL-W) and the Naval Reactors Facility. Excluded from the assessment were those HEU holdings previously assessed in the Idaho National Engineering Laboratory Spent Nuclear Fuel Inventory and Vulnerability Site Assessment Report and any HEU holdings evaluated in the Plutonium Vulnerability Assessment Report

  11. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    International Nuclear Information System (INIS)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S.

    2011-01-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched ( 235 U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  12. Automated instruments for in-line accounting of highly enriched uranium at the Oak Ridge Y-12 Plant

    International Nuclear Information System (INIS)

    Russo, P.A.; Strittmatter, R.B.; Sandford, E.L.; Stephens, M.M.; Brumfield, T.L.; Smith, S.E.; McCullough, E.E.; Jeter, I.W.; Bowers, G.L.

    1985-02-01

    Two automated nondestructive assay instruments developed at Los Alamos in support of nuclear materials accounting needs are currently operating in-line at the Oak Ridge Y-12 facility for recovery of highly enriched uranium (HEU). One instrument provides the HEU inventory in the secondary solvent extraction system, and the other monitors HEU concentration in the secondary intermediate evaporator. Both instruments were installed in December 1982. Operational evaluation of these instruments was a joint effort of Y-12 and Los Alamos personnel. This evaluation included comparison of the solvent extraction system inventories with direct measurements performed on the dumped solution components of the solvent extraction system and comparison of concentration assay results with the external assays of samples withdrawn from the process. The function and design of the instruments and detailed results of the operational evaluation are reported

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

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

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

  14. Communication received from the United Kingdom of Great Britain and Northern Ireland concerning its national holdings of civil high enriched uranium

    International Nuclear Information System (INIS)

    1998-01-01

    The document reproduces the text of the note verbale and its attachment dated 1 July 1998 received by the Director General of the IAEA from the Permanent Mission to the IAEA of the United Kingdom of Great Britain and Northern Ireland, making available information on its national holdings of civil high enriched uranium as of 31 December 1997

  15. Validation of COG10 and ENDFB6R7 on the Auk Workstation for General Application to Highly Enriched Uranium Systems

    Energy Technology Data Exchange (ETDEWEB)

    Percher, Catherine G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2011-08-08

    The COG 10 code package1 on the Auk workstation is now validated with the ENBFB6R7 neutron cross section library for general application to highly enriched uranium (HEU) systems by comparison of the calculated keffective to the expected keffective of several relevant experimental benchmarks. This validation is supplemental to the installation and verification of COG 10 on the Auk workstation2.

  16. The Text of an Agreement between the Agency and Spain Relating to the Application of Safeguards to a Quantity of Enriched Uranium

    International Nuclear Information System (INIS)

    1975-01-01

    The text of an agreement between the Agency and Spain relating to the application of safeguards to a quantity of enriched uranium is reproduced in this document for the information of all Members. The agreement entered into force on 18 June 1975, pursuant to Section 24.

  17. Communication received from the Resident Representative of the Russian Federation to the IAEA on the establishment, structure and operation of the International Uranium Enrichment Centre

    International Nuclear Information System (INIS)

    2007-01-01

    The Director General has received a communication dated 7 June 2007 from the Resident Representative of the Russian Federation, with an attachment entitled 'Establishment, Structure and Operation of the International Uranium Enrichment Centre'. As requested in that communication, the letter and its attachment are circulated for the information of Member States

  18. Environmental monitoring (operational period) of the uranium enrichment facility Almirante Alvaro Alberto. Quadrimonthly report of gamma spectroscopy measurements: march to june 1988

    International Nuclear Information System (INIS)

    Venturini, L.; Pecequilo, B.R.S.

    1990-02-01

    In this report we present the assessment of the environmental monitoring radiation levels during the operation period of the Uranium Enrichment Facility Almirante Alvaro Alberto from March to July 1988. The purpose was achieved by sampling and analyzing using gamma spectrometry, water and terrestrial and biological indicators. (author) [pt

  19. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  20. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    Primm, R.T. III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-01-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U 3 O 8 mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties

  1. 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

  2. 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

  3. Institutional arrangements for uranium enrichment U.S. contribution to INFCE working group 2

    International Nuclear Information System (INIS)

    1978-12-01

    In acquiring enrichment services, by contract or by facility construction, assurance of long term fuel supply and minimizing proliferation risks are two important factors. Various alternative institutional arrangements, ranging from construction of a national facility, reliance on a single foreign supplier or on several foreign sources, to multinational enrichment centers are reviewed. It is argued that for most nations the two specified goals are likely to be met by contracting with one or more foreign suppliers. The possible creation of an international fuel bank is explored briefly

  4. 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.

  5. An ideal cascade for uranium 235 enrichment by centrifuge jet nozzle process

    International Nuclear Information System (INIS)

    Santos, E.C. dos.

    1981-01-01

    The design of an ideal cascade for the process of isotope separation by centrifugation for the U 235 enrichment, is presented. A selection of building materials used in fabrication of isotope separation plants, showing the importance of aluminium, due the bauxite mines in Northern Brazil, is done. (M.C.K.) [pt

  6. Determination of D2O - 2% enriched uranium lattice parameters by means of a critical system

    International Nuclear Information System (INIS)

    Raisic, N.; Takac, S.; Markovic, H.; Bosevski, T.

    1963-01-01

    In order to specify experimental procedures for few standard measurements sufficient to provide consistent set of lattice parameters, a series of experiments were performed at the RB reactor using 2% enriched tubular fuel elements. Obtained results were compared to standard two-group diffusion calculation indicating high degree of accuracy for a broad variety of reactor lattice configurations

  7. Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant

    International Nuclear Information System (INIS)

    1996-08-01

    This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation

  8. Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation.

  9. Uranium market enters slow summer period as enrichment market heats up

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    Prices slipped in the restricted and unrestricted uranium markets. Prices in the restricted market remained above the $10-benchmark, with uranium trading in the relatively narrow range of $10.05 to $10.20 per lb U3O8. Meanwhile, NUKEM detected a lowering and widening of the range in the unrestricted market; $7.10 to $7.45 per lb U3O8 was the order of the day there. The other notable feature of recent price movements has been the aggressive competition in the US spot conversion market. Many analysts have noted favorable prices for UF6 relative to components in recent market activity. Summer doldrums appear to be setting in. May and a correspondingly small number of offers are outstanding as of the end of the month. Last month NUKEM inaugurated a spot conversion price range and asked for readers' views. Those expressing an opinion were unanimous; NUKEM will publish both its longstanding estimate of new contract prices as well as the spot range. Some noted the somewhat wide range of last month's spot conversion prices. In explanation, we note that this figure encompasses activity over the full month and in the US as well as the non-US markets. Thus, it resembles the spot SWU range

  10. H.R. 5181: A Bill to provide for the establishment of a uranium enrichment corporation, and for other purposes. Introduced in the House of Representatives, One Hundredth Congress, Second Session, August 8, 1988

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    H.R. 5181 is a bill to provide for the establishment of a uranium enrichmen corporation, which will be an efficient provider of uranium enrichment services able, insofar as is consistent with maximizing the value of the corporation, to meet the needs of the United States for uranium enrichment services, and which will be suitable for eventual transfer to private ownership. Other aspects of the bill assure the recovery of the value of assets transferred to the corporation, maximize the value of the corporation to the United States, assure the recovery of the costs for decommissioning existing and future uranium enrichment facilities owned by the corporation, and prevent any further public subsidization of the Civilian Uranium Enrichment Program in the United States. After introduction, the bill was referred jointly to the Committees on Energy and Commerce and Interior and Insular Affairs

  11. 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

  12. Environmental assessment for the purchase of Russian low enriched uranium derived from the dismantlement of nuclear weapons in the countries of the former Soviet Union

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    The United States is proposing to purchase from the Russian Federation low enriched uranium (LEU) derived from highly enriched uranium (HEU) resulting from the dismantlement of nuclear weapons in the countries of the former Soviet Union. The purchase would be accomplished through a proposed contract requiring the United States to purchase 15,250 metric tons (tonnes) of LEU (or 22,550 tonnes of UF{sub 6}) derived from blending 500 metric tones uranium (MTU) of HEU from nuclear warheads. The LEU would be in the form of uranium hexafluoride (UF{sub 6}) and would be converted from HEU in Russia. The United States Enrichment Corporation (USEC) is the entity proposing to undertake the contract for purchase, sale, and delivery of the LEU from the Russian Federation. The US Department of Energy (DOE) is negotiating the procedure for gaining confidence that the LEU is derived from HEU that is derived from dismantled nuclear weapons (referred to as ``transparency),`` and would administer the transparency measures for the contract. There are six environments that could potentially be affected by the proposed action; marine (ocean); US ports of entry; truck or rail transportation corridors; the Portsmouth GDP; the electric power industry; and the nuclear fuel cycle industry. These environmental impacts are discussed.

  13. Environmental assessment for the purchase of Russian low enriched uranium derived from the dismantlement of nuclear weapons in the countries of the former Soviet Union

    International Nuclear Information System (INIS)

    1994-01-01

    The United States is proposing to purchase from the Russian Federation low enriched uranium (LEU) derived from highly enriched uranium (HEU) resulting from the dismantlement of nuclear weapons in the countries of the former Soviet Union. The purchase would be accomplished through a proposed contract requiring the United States to purchase 15,250 metric tons (tonnes) of LEU (or 22,550 tonnes of UF 6 ) derived from blending 500 metric tones uranium (MTU) of HEU from nuclear warheads. The LEU would be in the form of uranium hexafluoride (UF 6 ) and would be converted from HEU in Russia. The United States Enrichment Corporation (USEC) is the entity proposing to undertake the contract for purchase, sale, and delivery of the LEU from the Russian Federation. The US Department of Energy (DOE) is negotiating the procedure for gaining confidence that the LEU is derived from HEU that is derived from dismantled nuclear weapons (referred to as ''transparency),'' and would administer the transparency measures for the contract. There are six environments that could potentially be affected by the proposed action; marine (ocean); US ports of entry; truck or rail transportation corridors; the Portsmouth GDP; the electric power industry; and the nuclear fuel cycle industry. These environmental impacts are discussed

  14. The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C; Diaz de la Rubia, T; Moses, E

    2008-12-23

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  15. Advances of the low enriched uranium utilization project in CNA-1 during 1998 and 1999

    International Nuclear Information System (INIS)

    Fink, Jose M.; Higa, Manabu; Sidelnik, Jorge I.; Perez, Ramon A.; Casario, Jose A.; Alvarez, Luis A.

    1999-01-01

    In this work, a general description of advances of the Enriched Fuel Introduction Project in CNA-1 and the main tasks performed during 1998 and 1999 are presented. The program is being satisfactorily developed and during that period the number of slightly enriched fuels (LEU) introduced had significantly increased in relation to previous years. At present, there are 181 LEU fuel elements in the core and 125 LEU fuel elements have been extracted. The number of full power burnt fuel elements per day decreased from 1.31 FE/dpp in 1994 (when all fuel was natural) to 0.92 in 1998 and 0.83 in 1999, reaching the predicted value for homogeneous LEU core of 0.7. The cost of burnt fuel in 1998 was 25% lower that if only natural fuel would have been used. (author)

  16. Nuclear criticality safety for warehousing of 55-gal drums containing highly enriched uranium metal

    International Nuclear Information System (INIS)

    Robinson, R.C.; Dodds, H.L.

    1989-01-01

    Subcritical data for six separate arrays of 55-gal 17H steel shipping/storage containers is calculated and used to determine safe storage limits for U(97.5)-metal in a warehouse environment at the Oak Ridge Y-12 Plant. Results are presented for three different forms of fissile material: 97.5 wt% 235 U enriched metal (ρ 0 = 18.76 g/cm 3 ) cylinders, metal spheres, and low density (ρ = 0.25 ρ 0 ) metal cylinders

  17. Other enrichment related contracts

    International Nuclear Information System (INIS)

    Hall, J.C.

    1978-01-01

    In addition to long-term enrichment contracts, DOE has other types of contracts: (1) short-term, fixed-commitment enrichment contract; (2) emergency sales agreement for enriched uranium; (3) feed material lease agreement; (4) enriched uranium storage agreement; and (5) feed material usage agreement

  18. Economic and Non-proliferation Policy Considerations of Uranium Enrichment in Brazil and Argentina

    International Nuclear Information System (INIS)

    Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.; Mahy, Heidi A.

    2008-01-01

    The nuclear development programs of both Argentina and Brazil have, since the 1970s, been premised on the desire for self-sufficiency and assurance of nuclear fuel supply. While military rivalry and mutual distrust led to nuclear weapons related development programs in the 1970s and 1980s, both countries have since terminated these programs. Furthermore, the governments of both countries have pledged their commitment to exclusively non-explosive use of nuclear energy and have signed the Non Proliferation Treaty (NPT). Utilizing rights provided for under the NPT, both Argentina and Brazil have nuclear fuel production facilities, with the notable exception of enrichment plants, that provide much of the current indigenous fuel requirements for their nuclear power plants. However, both countries are actively developing enrichment capability to fill this gap. The purpose of this report is to assess the economic basis and non-proliferation policy considerations for indigenous enrichment capability within the context of their desired self-sufficiency and to evaluate possible United States Government policy options.

  19. Economic and Non-proliferation Policy Considerations of Uranium Enrichment in Brazil and Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.; Mahy, Heidi A.

    2008-09-01

    The nuclear development programs of both Argentina and Brazil have, since the 1970s, been premised on the desire for self-sufficiency and assurance of nuclear fuel supply. While military rivalry and mutual distrust led to nuclear weapons related development programs in the 1970s and 1980s, both countries have since terminated these programs. Furthermore, the governments of both countries have pledged their commitment to exclusively non-explosive use of nuclear energy and have signed the Non Proliferation Treaty (NPT). Utilizing rights provided for under the NPT, both Argentina and Brazil have nuclear fuel production facilities, with the notable exception of enrichment plants, that provide much of the current indigenous fuel requirements for their nuclear power plants. However, both countries are actively developing enrichment capability to fill this gap. The purpose of this report is to assess the economic basis and non-proliferation policy considerations for indigenous enrichment capability within the context of their desired self-sufficiency and to evaluate possible United States Government policy options.

  20. Method for pyrohydrolitic decomposition of phosphorus-containing liquid contaminated with highly enriched uranium

    International Nuclear Information System (INIS)

    Vietzke, H.; Hofmann, J.; Wagner, G.

    1984-01-01

    Wastes containing haloorganics and/or phosphorus compounds are incinerated in a steam-fluidized bed of Al and/or Mg and/or Ca compounds, e.g. CaO, at 300-800.degree.; the waste gases are then incinerated at approx. 1000.degree. and scrubbed at pH 5. A 3/7 tri-Bu phosphate kerosene mixture contaminated with Uranium was sprayed into the bed at 15 kg/h. The afterburner at 100.degree. was fed with 33 m3 oxygen/h to facilitate complete combustion to CO2 and H2O. After 36 h the CaO was exhausted. The recirculating waters in the scrubber contained only 0.02% of the activity of the wastes and the gaseous effluent none activity. (orig./PW)

  1. S.210: A Bill to establish the United States Enrichment Corporation to operate the Federal uranium enrichment program on a profitable and efficient basis in order to maximize the long term economic value to the United States, to provide assistance to the domestic uranium industry and to provide a Federal contribution for the reclamation of mill tailings generated pursuant to Federal defense contracts at active uranium and thorium processing sites, introduced in the Senate of the United States, One Hundred Second Congress, First Session, January 15, 1991

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    The bill would amend the Atomic Energy Act of 1954 to establish the US Enrichment Corporation to operate the Federal uranium enrichment program on a profitable and efficient basis in order to maximize the long term economic value to the US, would provide a Federal contribution for the reclamation of mill tailings generated as a result of Federal defense contracts at active uranium and thorium processing sites, and would provide assistance to the domestic uranium industry

  2. Uranium industry annual 1996

    International Nuclear Information System (INIS)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs

  3. Uranium industry annual 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  4. Petrochemical and Tectonogenesis of Granitoids in the Wuyo-Gubrunde Horst, Northeastern Nigeria: Implication for Uranium Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Bolarinwa, Anthony Temidayo, E-mail: atbola@yahoo.com; Bute, Saleh Ibrahim [University of Ibadan, Department of Geology (Nigeria)

    2016-06-15

    The Wuyo-Gubrunde Horst in the northeastern Nigeria consists of migmatite gneiss, unaltered, altered, and sheared porphyritic granites, pegmatites, aplites, basalts, and sandstone. Uranium has been reported in rhyolite, sheared rocks, and sandstone within the area. The petrogenesis of the granitoids and associated rocks in the area was evaluated in the light of new geochemical data, which showed that the U content of altered porphyritic granite is highest and hydrothermal-related. The granitoids are metaluminous, sub-alkaline, and S-type granite, and have evolved by partial melting of crustal material emplaced at moderate depth of 20–30 km in a syn-to late-collisional within-plate tectonic setting. The negative Eu/Eu* anomaly and high (La/Yb){sub N} ratio of the granitoids indicate magma fractionation. The low SiO{sub 2} (<53%) and high Fe{sub 2}O{sub 3} (10%) of the altered porphyritic granite compared to other similar rock units suggest pervasive alteration. The associated basalts are tholeiitic, emplaced within continental plate tectonic setting, and enriched in Ni, V, Nb, Sr, and light rare earth elements, and they have SiO{sub 2}, Fe{sub 2}O{sub 3}, V, Th, and Co contents that are similar to those of the altered porphyritic granites. The U occurrence in the Wuyo-Gubrunde Horst is believed to be sourced from the adjoining Bima sandstone in the Benue Trough, which locally contains carbonaceous zones with anomalously high concentrations of U. The Fe{sup 2+}/Fe{sup 3+} redox fronts formed by alteration of the iron-rich basalts provided the requisite geochemical barrier for U-bearing hydrothermal fluid, causing enrichment of U leached and mobilized from the sandstone through fractures in the rocks.

  5. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  6. Feasibility studies to establish at the Kazakhstan Ulba metallurgical plant the manufacturing capability to produce low-enriched uranium certified reference materials

    Energy Technology Data Exchange (ETDEWEB)

    Kuzminski, Jozef [Los Alamos National Laboratory; Nesuhoff, J [NBL; Cratto, P [NBL; Pfennigwerth, G [Y12 NATIONAL SEC. COMPLEX; Mikhailenko, A [ULBA METALLURGICAL PLANT; Maliutina, I [ULBA METALLURGICAL PLANT; Nations, J [GREGG PROTECTION SERVICES

    2009-01-01

    One of the salient features of the transition plan that the United States Department of Energy/National Nuclear Security Administration (DOE/NNSA) is presently implementing in the Former Soviet Union countries is the availability of uranium certified reference materials for calibration of nondestructive assay (NDA) measurement equipment. To address this challenge, DOE/NNSA and U.S. national laboratories have focused their cooperative efforts on establishing a reliable source for manufacturing, certifying, and supplying of such standards. The Ulba Metallurgical Plant (UMP), Kazakhstan, which processes large quantities of low-enriched uranium to produce ceramic fuel pellets for nuclear-powered reactors, is well situated to become a key supplier of low-enriched uranium certified reference materials for the country and Central Asia region. We have recently completed Phase I of a feasibility study to establish at UMP capabilities of manufacturing these standards. In this paper we will discuss details of a proposed methodology for uranium down-blending, material selection and characterization, and a proposed methodology of measurement by destructive (DA) and non-destructive (NDA) analysis to form a database for material certification by the competent State authorities in the Republic of Kazakhstan. In addition, we will discuss the prospect for manufacturing of such standards at UMP.

  7. Department of Energy`s Uranium Enrichment Decontamination and Decommissioning Fund fiscal year 1997 financial statement audit

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-08-21

    This report presents the results of the independent certified public accountants` audit of the Department of Energy`s Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) financial statements as of September 30, 1997. The auditors have expressed an unqualified opinion on the 1997 statement of financial position and the related statements of operations and changes in net position and cash flows. The 1997 financial statement audit was made under provisions of the Inspector General Act (5 U.S.C. App.) as amended, the Government Management Reform Act (31 U.S.C. 3515), and Office of Management and Budget implementing guidance. The auditor`s work was conducted in accordance with generally accepted government auditing standards. To fulfill our audit responsibilities, we contracted with the independent public accounting firm of KPMG Peat Marwick LLP (KPMG) to conduct the audit for us, subject to our review. The auditors` report on the D&D Fund`s internal control structure disclosed no reportable conditions. The auditors` report on compliance with laws and regulations disclosed one instance of noncompliance. This instance of noncompliance relates to the shortfall in Government appropriations. Since this instance was addressed in a previous audit, no further recommendation is made at this time. During the course of the audit, KPMG also identified other matters that, although not material to the financial statements, nevertheless, warrant management`s attention. These items are fully discussed in a separate letter to management.

  8. The mortality and cancer morbidity experience of workers at the Capenhurst uranium enrichment facility 1946-95

    International Nuclear Information System (INIS)

    McGeoghegan, D.

    2000-01-01

    The results presented here contain the follow-up of the cohort of workers ever employed at the Capenhurst site of British Nuclear Fuels plc or its predecessors between 1946 and 1995. The main activity of the plant is isotopic, 235 U, enrichment of uranium. The study cohort consists of 12,540 employees and contains 334,473 person-years of follow up. This is a relatively mature cohort, with a mean follow-up period of 26.7 years, that has been exposed to low levels of radiation. The collective external radiation dose received by the 3244 radiation workers was 31.95 person-sieverts, with mean cumulative dose 9.85 mSv. To the end of 1995 there have been 3841 deaths recorded for this cohort, 585 of which were amongst radiation workers. The standardised mortality ratios (SMRs) for all causes were significantly low, 83 and 91 respectively, for radiation and non-radiation workers, indicating the usual 'healthy worker' effect. The cancer mortality was less than that expected, though not significantly so, with SMRs for all cancers of 88 and 97, for radiation and non-radiation workers respectively. The cancer registration rates were significantly low, with standardised registration ratios (SRRs) for all cancers of 82 and 88, for radiation and non-radiation workers respectively. An association between bladder cancer registrations and cumulative external radiation exposure was noted when the cumulative external dose was lagged by 20 years. (author)

  9. The mortality and cancer morbidity experience of workers at the Capenhurst uranium enrichment facility 1946-95

    Energy Technology Data Exchange (ETDEWEB)

    McGeoghegan, D. [Westlakes Scientific Consulting Ltd, The Princess Royal Building, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3LN (United Kingdom). E-mail: david.mcgeoghegan at westlakes.ac.uk; Binks, K. [Westlakes Scientific Consulting Ltd, The Princess Royal Building, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3LN (United Kingdom)

    2000-12-01

    The results presented here contain the follow-up of the cohort of workers ever employed at the Capenhurst site of British Nuclear Fuels plc or its predecessors between 1946 and 1995. The main activity of the plant is isotopic, {sup 235}U, enrichment of uranium. The study cohort consists of 12,540 employees and contains 334,473 person-years of follow up. This is a relatively mature cohort, with a mean follow-up period of 26.7 years, that has been exposed to low levels of radiation. The collective external radiation dose received by the 3244 radiation workers was 31.95 person-sieverts, with mean cumulative dose 9.85 mSv. To the end of 1995 there have been 3841 deaths recorded for this cohort, 585 of which were amongst radiation workers. The standardised mortality ratios (SMRs) for all causes were significantly low, 83 and 91 respectively, for radiation and non-radiation workers, indicating the usual 'healthy worker' effect. The cancer mortality was less than that expected, though not significantly so, with SMRs for all cancers of 88 and 97, for radiation and non-radiation workers respectively. The cancer registration rates were significantly low, with standardised registration ratios (SRRs) for all cancers of 82 and 88, for radiation and non-radiation workers respectively. An association between bladder cancer registrations and cumulative external radiation exposure was noted when the cumulative external dose was lagged by 20 years. (author)

  10. Communication received from the Governor of Norway to the Agency concerning the International Symposium of Minimisation of Highly Enriched Uranium in the Civilian Nuclear Sector

    International Nuclear Information System (INIS)

    2006-01-01

    The Director General has received a communication from the Governor of Norway, attaching the Chair's Summary of the discussions held during the International Symposium of Minimisation of Highly Enriched Uranium in the Civilian Nuclear Sector which was held in Oslo from 17 to 20 June 2006 as well as the summary from the technical workshop of the Symposium. The communication and, as requested therein, the attached two summaries, are herewith circulated for the information of Member States

  11. Finding of no significant impact: Interim storage of enriched uranium above the maximum historical level at the Y-12 Plant Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1995-01-01

    The US Department of Energy (DOE) has prepared an Environmental Assessment (EA) for the Proposed Interim Storage of Enriched Uranium Above the Maximum Historical Storage Level at the Y-12 Plant, Oak Ridge, Tennessee (DOE/EA-0929, September, 1994). The EA evaluates the environmental effects of transportation, prestorage processing, and interim storage of bounding quantities of enriched uranium at the Y-12 Plant over a ten-year period. The State of Tennessee and the public participated in public meetings and workshops which were held after a predecisional draft EA was released in February 1994, and after the revised pre-approval EA was issued in September 1994. Comments provided by the State and public have been carefully considered by the Department. As a result of this public process, the Department has determined that the Y-12 Plant-would store no more than 500 metric tons of highly enriched uranium (HEU) and no more than 6 metric tons of low enriched uranium (LEU). The bounding storage quantities analyzed in the pre-approval EA are 500 metric tons of HEU and 7,105.9 metric tons of LEU. Based on-the analyses in the EA, as revised by the attachment to the Finding of No Significant Impact (FONSI), DOE has determined that interim storage of 500 metric tons of HEU and 6 metric tons of LEU at the Y-12 Plant does not constitute a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement (EIS) is not required and the Department is issuing this FONSI

  12. Australia's uranium

    International Nuclear Information System (INIS)

    Hampson, D.C.

    1980-01-01

    The subject is discussed as follows: structure of the uranium industry in Australia (export policies; development of mining programme; table of export contracts approved by Australian government, 1972; government policy towards the industry 1972-75 and since 1975); reserves (table of Australia's major uranium deposits; estimated world resources of uranium, excluding USSR, Eastern Europe and China; comparison of exploration expenditures and discovery of uranium in Australia and the USA); enrichment; resource potential; future demand (table of nuclear power reactors above 30 MW in operation or under construction, mid-1979; projection of Australian uranium production to 1990); government and union action. (U.K.)

  13. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  14. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  15. Additional supply agreement of 9 December 1988 between the International Atomic Energy Agency and the Governments of the Islamic Republic of Iran and of Argentina for the transfer of enriched uranium for a research reactor in Iran

    International Nuclear Information System (INIS)

    1990-01-01

    The document reproduces the text of the additional supply agreement of 9 December 1988 between the IAEA and the Government of the Islamic Republic of Iran and of Argentina for the transfer of 115.80 kilograms of uranium enriched up to 20% in the isotope uranium-235 for a research reactor in Iran

  16. Research on the conversion of highly enriched uranium (HEU) nitrate by using the microwave denitration

    International Nuclear Information System (INIS)

    Bao Weimin; Song Chongli

    1998-08-01

    In order to simplify the denitration process by microwave heating, the uranyl nitrate is firstly denitrated and converted into UO 3 . The produced UO 3 was then further heated in the microwave field to transfer UO 3 to U 3 O 8 and to form a single product of U 3 O 8 . When the phase transfer from UO 3 to U 3 O 8 occurs, the temperature of the product increases 200∼300 degree C in two minutes. The phase-transfer temperature can be controlled by the input power of microwave. High quality U 3 O 8 can be obtained at a denitration temperature about 500 degree C. It contains no residual NO x and has a specific surface area great than 3 m 2 /g. The denitration temperature is measured with an IR-thermometer and checked with an optic fiber thermometer. The working curve and process parameter were studied in a microwave denitration unit for high enriched uranyl nitrate solution (90 g(U)/L, 4 mol/L HNO 3 and 1.2 L per batch)

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

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

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

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

    International Nuclear Information System (INIS)

    SCHWINKENDORF, K.N.

    2006-01-01

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

  19. Uranium industry annual, 1991

    International Nuclear Information System (INIS)

    1992-10-01

    In the Uranium Industry Annual 1991, data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2. A feature article entitled ''The Uranium Industry of the Commonwealth of Independent States'' is included in this report

  20. S. 2415: Title I may be cited as the Uranium Enrichment Act of 1990; Title II may be cited as the Uranium Security and Tailings Reclamation Act of 1989; and Title III may be cited as The Solar, Wind, Waste, and Geothermal Power Production Incentives Act of 1990, introduced in the Senate, One Hundred First Congress, Second Session, April 4, 1990

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    S. 2415 (which started out as a bill to encourage solar and geothermal power generation) now would amend the Atomic Energy Act of 1954 to redirect uranium enrichment enterprises to further the national interest, respond to competitive market forces, and to ensure the nation's common defense and security. It would establish a United States Enrichment Corporation for the following purposes: to acquire feed materials, enriched uranium, and enrichment facilities; to operate these facilities; to market enriched uranium for governmental purposes and qualified domestic and foreign persons; to conduct research into uranium enrichment; and to operate as a profitable, self-financing, reliable corporation and in a manner consistent with the health and safety of the public. The bill describes powers and duties of the corporation; the organization, finance, and management; decontamination and decommissioning. The second part of the bill would ensure an adequate supply of domestic uranium for defense and power production; provide assistance to the domestic uranium industry; and establish, facilitate, and expedite a comprehensive system for financing reclamation and remedial action at active uranium and thorium processing sites. The third part of the bill would remove the size limitations on power production facilities now part of the Public Utility Regulatory Policies Act of 1978. Solar, wind, waste, or geothermal power facilities would no longer have to be less than 80 MW to qualify as a small power production facility

  1. Feasibility and options for purchasing nuclear weapons, highly enriched uranium (HEU) and plutonium from the former Soviet Union (FSU)

    International Nuclear Information System (INIS)

    1994-01-01

    In response to a recent tasking from the National Security Council, this report seeks to analyze the possible options open to the US for purchasing, from the former Soviet Union (FSU) substantial quantities of plutonium and highly enriched uranium recovered from the accelerated weapons retirements and dismantlements that will soon be taking place. The purpose of this paper is to identify and assess the implications of some of the options that now appear to be open to the United States, it being recognized that several issues might have to be addressed in further detail if the US Government, on its own, or acting with others seeks to negotiate any such purchases on an early basis. As an outgrowth of the dissolution of the Soviet Union three of the C.I.S. republics now possessing nuclear weapons, namely the Ukraine, Belarus, and Kazakhstan, have stated that it is their goal, without undue delay, to become non-nuclear weapon states as defined in the Non-Proliferation Treaty. Of overriding US concern is the proliferation of nuclear weapons in the Third World, and the significant opportunity that the availability of such a large quantity of surplus weapons grade material might present in this regard, especially to a cash-starved FSU Republic. Additionally, the US, in its endeavor to drawdown its own arsenal, needs to assure itself that these materials are not being reconfigured into more modern weapons within the CIS in a manner which would be inconsistent with the stated intentions and publicized activities. The direct purchase of these valuable materials by the US government or by interested US private enterprises could alleviate these security concerns in a straightforward and very expeditious manner, while at the same time pumping vitally needed hard currency into the struggling CIS economy. Such a purchase would seem to be entirely consistent with the Congressional mandate indicated by the Soviet Nuclear Threat Reduction Act of 1991

  2. Environmental Assessment for the shipment of low enriched uranium billets to the United Kingdom from the Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1992-08-01

    This Environmental Assessment provides the necessary information so that a decision can be made on whether a Finding of No Significant Impact Environmental Impact Statement should be prepared for the proposed action. The proposed action is to transfer 2,592 low enriched uranium billets to the United Kingdom. The billets are currently stored in the 300 Area of the Hanford Site, Richland, Washington. The proposed action would consist of two types of activities: loading and transportation. The loading activities would include placing the billets into the appropriate containers for transportation. The transportation activities would include the tasks required to transport the containers 215 miles (344 km) via highway to the Port of Seattle, Washington, and transfer the containers aboard an ocean cargo vessel for transportation to the United Kingdom. The Department of Energy would only be responsible for conducting the loading activities. The United Kingdom would be responsible for conducting the transportation activities in compliance with all applicable United States and international transportation laws. The tasks associated with the proposed action activities have been performed before and are well defined in terms of requirements and consequences. A risk assessment and a nuclear safety evaluation were performed to address safety issues associated with the proposed action. The risk assessment determined the exposure risk from normal operation and from the maximum credible accident that involves a truck or ship collision followed by a fire that engulfs all the billets in the shipment and the release of the radiological contents of the shipment to the environment. The criticality assessment determined the nuclear safety limits for handling, transporting and storing the shipment under incident-free and accident transport conditions

  3. Highly enriched uranium (HEU) politics: An enigma wrapped up in a warhead and boxed in political chaos

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    It could be fairly said that while the Cold War arose in an atmosphere of mutual mistrust and hostility, it is ending with an equal amount of confusion and uncertainty. More than a year has passed since the US and Russia signed a tentative HEU agreement in August 1992. Many of the details have been worked out, but major questions remain. And they're not just on the Russian side. The fine points of President Clinton's overall nuclear policy are only now beginning to emerge. In his first major foreign policy address, before the United Nations in late September, Clinton called for a worldwide ban on the production of plutonium and HEU for nuclear weapons. open-quotes Growing global stockpiles of plutonium and highly enriched uranium are raising the danger of nuclear terrorism for all nations,close quotes said Clinton before the UN. open-quotes We will press for an international agreement that would ban production of these materials for weapons forever.close quotes As the veil lifts from Clinton's nuclear policy, it appears the Administration realizes that Russia may have more HEU than originally thought. That possibility has been confirmed by Minatom Minister Mikhailov's disclosures to the NUKEM Market Report, which brought a greater degree of certainty to estimates that had been floating around for some time. When the Bush Administration signed the HEU pact, it apparently thought the 500 metric tons comprised most of the former Soviet Union's nuclear arsenal. Now that the number appears higher, Clinton may propose to accelerate and enlarge the HEU deal. He is due to summit with Yeltsin, if Yeltsin survives, next spring. The 500-metric-ton deal may only be the first step

  4. Steady-state thermal hydraulic analysis of Tajoura Nuclear Research Centre reactor (TNRC) with low enriched uranium core

    International Nuclear Information System (INIS)

    Ghangir, F.; Elwaer, A.

    2015-01-01

    This study focuses on the steady-state thermal hydraulic analysis of the reactor of the Tajoura Nuclear Research Centre (TNRC) with LOW Enriched Uranium (LEU) core. A MATLAB program has been constructed to calculate the important related safety parameters at the steady-state operation of the core. The calculations have been performed at the hottest cell of the core and at the operating power of 10 MW with coolant inlet temperature equals to 45 °. The benefit of the symmetry of the shape of the core cells has been used to divide the hottest cell to four regions. Each quarter of the hottest cell has been further divided to a plate part and a corner part. Therefore, the calculations have been applied to the quarter of the hottest cell at both plate and corner parts. The assessment of the operating power according to the maximum allowable clad surface temperature is used to identify the maximum operating power of TNRC's reactor. The outcome of this investigation is that the reactor of the TNRC must not operate at power equals to 10 MW and the maximum allowable operating power is 9.7 MW where the value of the maximum clad surface temperature is 101.7 ° (maximum allowable clad surface temperature = 102 °). The results have been compared witha previous studies [1] where a Nusselt number is calculated using a different correlation equation from this study. [2] where an onset of nucleate boiling (ONB) and the maximum allowable power are tested and with the results of a team at Argonne National Lab, ANL, [3] Which present a good agreement.(author)

  5. The pressure vessel type heavy water reactor on the natural and slightly enriched uranium once-through fuel cycle

    International Nuclear Information System (INIS)

    1979-06-01

    This paper deals with the pressurized heavy water reactor which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources

  6. United States Department of Energy, Office of Environmental Management, Uranium Enrichment Decontamination and Decomissioning Fund financial statements, September 30, 1996 and 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-05-01

    The Energy Policy Act of 1992 (Act) established the Uranium Enrichment Decontamination and Decommissioning Fund (D and D Fund, or Fund) to pay the costs for decontamination and decommissioning three gaseous diffusion facilities located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio (diffusion facilities). The Act also authorized the Fund to pay remedial action costs associated with the Government`s operation of the facilities and to reimburse uranium and thorium licensees for the costs of decontamination, decommissioning, reclamation, and other remedial actions which are incident to sales to the Government. The report presents the results of the independent certified public accountants` audit of the D and D Fund financial statements as of September 30, 1996. The auditors have expressed an unqualified opinion on the 1996 statement of financial position and the related statements of operations and changes in net position and cash flows.

  7. Comparison of low enriched uranium (UAlx-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

    International Nuclear Information System (INIS)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G.; Nishiyama, Pedro J.B. de O.

    2011-01-01

    The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium ( 235 U) UAl x dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

  8. Uranium enrichment using laser and plasma separation - an analysis focussing on export control; Urananrikning med laser och plasmaseparation - en analys med fokus paa exportkontroll

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Wilhelmsen, Katarina; Wirstam, Jens (Swedish Defence Research Agency, FOI, Stockholm (SE))

    2007-07-01

    The Swedish Defence Research Agency, FOI, has under contract work financed by the Swedish Nuclear Power Inspectorate, SKI, performed a study on uranium enrichment by laser-based processes and plasma separation. None of these processes are at present used on an industrial scale for uranium enrichment. However, these are processes of high efficiency which implicates that plants employing any of these processes for production of fissile material for nuclear weapons can be small in size and easily hidden. This has also been the case for laser enrichment. Special emphasis has been put on the understanding of the governing physical principles of the different processes. In the report the AVLIS and MLIS processes are described in some detail while CRISLA and SILEX are treated in less detail. The plasma separation process is described in a separate section. Limiting physical parameters for the different processes are discussed and equipment relevant to the different technologies is described. Further, some indicators of the different processes are identified and listed. In the final chapter known research programs where these processes have been used are listed. Since none of the described technologies has been used on an industrial scale, information on plant design and running is scarce

  9. Price of military uranium

    International Nuclear Information System (INIS)

    Klimenko, A.V.

    1998-01-01

    The theoretical results about optimum strategy of use of military uranium confirmed by systems approach accounts are received. The numerical value of the system approach price of the highly enriched military uranium also is given

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

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

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

  11. Uranium industry annual, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Uranium industry data collected in the EIA-858 survey provide a comprehensive statistical characterization of annual activities of the industry and include some information about industry plans over the next several years. This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities utility market requirements and related topics

  12. Update on Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium Oxide

    International Nuclear Information System (INIS)

    Mount, M.; O'Connell, W.; Cochran, C.; Rinard, P.; Dearborn, D.; Endres, E.

    2002-01-01

    In October of 1999, Lawrence Livermore National Laboratory (LLNL) began an effort to calibrate the LLNL passive-active neutron (PAN) drum shuffler for measurement of highly enriched uranium (HEU) oxide. A single unit of certified reference material (CRM) 149 (Uranium (93% Enriched) Oxide - U 3 O 8 Standard for Neutron Counting Measurements) was used to (1) develop a mass calibration curve for HEU oxide in the nominal range of 393 g to 3144 g 235 U, and (2) perform a detailed axial and radial mapping of the detector response over a wide region of the PAN shuffler counting chamber. Results from these efforts were reported at the Institute of Nuclear Materials Management 4lSt Annual Meeting in July 2000. This paper describes subsequent efforts by LLNL to use a unit of CRM 146 (Uranium Isotopic Standard for Gamma Spectrometry Measurements) in consort with Monte Carlo simulations of the PAN shuffler response to CRM 149 and CRM 146 units and a selected set of containers with CRM 149-equivalent U 3 O 8 to (1) extend the low range of the reported mass calibration curve to 10 g 235 U, (2) evaluate the effect of U 3 O 8 density (2.4 g/cm 3 to 4.8 g/cm 3 ) and container size (5.24 cm to 12.17 cm inside diameter and 6.35 cm to 17.72 cm inside height) on the PAN shuffler response, and (3) develop mass calibration curves for U 3 O 8 enriched to 20.1 wt% 235 U and 52.5 wt% 235 U.

  13. Uranium Enrichment Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Demuth, Scott F. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-26

    DIV of facility layout, material flows, and other information provided in the DIQ. Material accountancy through an annual PIV and a number of interim inventory verifications, including UF6 cylinder identification and counting, NDA of cylinders, and DA on a sample collection of UF6. Application of C/S technologies utilizing seals and tamper-indicating devices (TIDs) on cylinders, containers, storage rooms, and IAEA instrumentation to provide continuity of knowledge between inspection. Verification of the absence of undeclared material and operations, especially HEU production, through SNRIs, LFUA of cascade halls, and environmental swipe sampling

  14. 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

  15. Nuclear threat initiative - Kazakhstan project on elimination of high-enriched uranium. 8 October 2005, Ust Kamenogorsk, Kazakhstan

    International Nuclear Information System (INIS)

    ElBaradei, M.

    2005-01-01

    Since its establishment in 2001, NTI (Nuclear Threat Initiative) has made important contributions towards securing weapon-usable nuclear material and reducing the threat of nuclear terrorism. Shortly after the terrorist attacks of September 2001, NTI pledged $1.15 million to the Agency?s Nuclear Security Fund, which has been used to achieve tangible nuclear security improvements: upgrades to physical protection of nuclear facilities and nuclear and radioactive material; urgently needed training in nuclear security for national officials; enhanced detection capabilities at border crossings; and improved national and international readiness for responding to terrorist acts. NTI has consistently sought to target its contributions to address areas of high priority. A primary point of focus has been to secure (and, where possible, eliminate) material that could be diverted for weapons purposes. In P roject Vinca , NTI committed $5 million in 2002 to help remove high enriched uranium fuel from a research reactor near Belgrade, for return to Russia. With support from Russia, the United States and NTI, seven transfers of fresh fuel back to Russia have been made since 2002 - a total of 112 kilograms of HEU. Current plans foresee further shipments of fresh HEU from another three countries in the next 15 months. The IAEA is also continuing to work on arrangements for the repatriation of spent research reactor fuel of Russian origin. In 2001, Sam Nunn and NTI president Charles Curtis approached the Kazakh Government to offer support for the safe transportation of the unused nuclear fuel from the shutdown BN-350 reactor - nearly 3000 kilograms - to the Ulba Metallurgical Plant JSC (UMP), where it could be dismantled and down-blended into LEU (NTI-Kazakhstan project). By the end of this year, 2897 kilograms of HEU - enough to produce dozens of nuclear bombs - will have been down-blended to LEU and placed in safe storage. Throughout the project, the IAEA has been implementing

  16. Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

    Directory of Open Access Journals (Sweden)

    Jaroszewicz Janusz

    2014-07-01

    Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

  17. Parametric study of the low-enriched uranium integrated Fort-Saint-Vrain element; comparative evaluation with the interacting tubular element

    International Nuclear Information System (INIS)

    Cerles, J.M.; Carvallo, G.; Vallepin, C.

    1971-11-01

    This paper presents a study of the influence of the different geometric and neutronic parameters on the calculation of the cycle with low-enriched uranium in a Fort-Saint-Vrain type brick. The study is divided in two parts: a stage of physics, essentially neutronics; an economical part where the costs are taken into account. At the level of studies of neutronics and costs, a parallel comparison is developed between the brick Fort-Saint-Vrain and the interacting tubular element, and even thorium. 6 refs. 29 figs [fr

  18. Provision by the uranium and uranium products

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2005-01-01

    International uranium market is converted from the buyer market into the seller market. The prices of uranium are high and the market attempts to adapt to changing circumstances. The industry of uranium enrichment satisfies the increasing demands but should to increase ots capacities. On the whole the situation is not stable and every year may change the existing position [ru

  19. Preliminary analyses of critical experiments using 10% enriched uranium nitrate solution with the 800mm-diameter cylindrical tank of STACY

    Energy Technology Data Exchange (ETDEWEB)

    Sono, Hiroki; Miyoshi, Yoshinori; Ohno, Akio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-05-01

    STACY, the static experiment critical facility in NUCEF, has the plan of a series of critical experiments using 10% enriched uranium nitrate solution with the 800mm-diameter cylindrical tank. The main purpose of these experiments is to measure the critical solution height and the temperature coefficient of reactivity by varying concentration of uranium and temperature of the fuel solution. Prior to these experiments, it is needed to make a diluting program of the solution, which is utilized to adjust the concentration of uranium, and to estimate the reactivity effect caused by temperature change. Therefore, such neutronic characteristics as critical solution height, height differential reactivity and kinetics parameters, were calculated with computational codes. In these preliminary analyses, simplified formulas to evaluate the above neutronic characteristics were also obtained by the least squares method applied to the results of computations. These formula well agree with computations within 0.1% up to 3.5%, so that they are useful to make detailed plans for the experiments and the reactor operation. Then, the temperature coefficient of reactivity is estimated at 3.85cent/degC, approximately. (author)

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

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

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

  1. Experiments and Simulations of the Use of Time-Correlated Thermal Neutron Counting to Determine the Multiplication of an Assembly of Highly Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    David L. Chichester; Mathew T. Kinlaw; Scott M. Watson; Jeffrey M. Kalter; Eric C. Miller; William A. Noonan

    2014-11-01

    A series of experiments and numerical simulations using thermal-neutron time-correlated measurements has been performed to determine the neutron multiplication, M, of assemblies of highly enriched uranium available at Idaho National Laboratory. The experiments used up to 14.4 kg of highly-enriched uranium, including bare assemblies and assemblies reflected with high-density polyethylene, carbon steel, and tungsten. A small 252Cf source was used to initiate fission chains within the assembly. Both the experiments and the simulations used 6-channel and 8-channel detector systems, each consisting of 3He proportional counters moderated with polyethylene; data was recorded in list mode for analysis. 'True' multiplication values for each assembly were empirically derived using basic neutron production and loss values determined through simulation. A total of one-hundred and sixteen separate measurements were performed using fifty-seven unique measurement scenarios, the multiplication varied from 1.75 to 10.90. This paper presents the results of these comparisons and discusses differences among the various cases.

  2. 235U isotopic characterization of natural and enriched uranium materials by using multigroup analysis (MGA method at a defined geometry using different absorbers and collimators

    Directory of Open Access Journals (Sweden)

    Yücel Haluk

    2015-09-01

    Full Text Available Characterization of nuclear materials is an important topic within the context of nuclear safeguards, homeland security and nuclear forensics. This paper deals with the performance of multigroup gamma-ray analysis (MGA method using the X- and γ-rays in the 80-130 keV region and enrichment meter principle (EMP based on the analysis of 185.7 keV peak for a certain geometry using different absorbers and collimators. The results from MGA and those of EMP are compared. In particular, the effect of aluminum/lead absorbers and lead collimator on the enrichment determination of 235U in natural and low enriched samples is investigated in a given source-detector geometry. The optimum diameter/height ratio for the Pb-collimator is found to be Dc/Hc = 1.4-1.6 in the chosen geometry. In order to simulate the container walls, ten different thicknesses of Al-absorbers of 141 to 840 mg·cm-2 and six different thicknesses of Pb-absorbers of 1120-7367 mg·cm-2 are interposed between sample and detector. The calibration coefficients (% enrichment/cps are calculated for each geometry. The comparison of the MGA and EMP methods shows that the enrichment meter principle provides more accurate and precise results for 235U abundance than those of MGA method at the chosen geometrical conditions. The present results suggest that a two-step procedure should be used in analyses of uranium enrichment. Firstly MGA method can be applied in situ and then EMP method can be used at a defined geometry in laboratory.

  3. Uranium Industry Annual, 1992

    International Nuclear Information System (INIS)

    1993-01-01

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ''Decommissioning of US Conventional Uranium Production Centers,'' is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2

  4. Uranium Industry Annual, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

  5. Uranium management activities

    International Nuclear Information System (INIS)

    Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

    2001-01-01

    One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

  6. Project and supply agreement. The text of the agreement of 20 September 1990 between the International Atomic Energy Agency and the Governments of the Republic of Zaire and the United States of America concerning the transfer of enriched uranium for a research reactor

    International Nuclear Information System (INIS)

    1991-01-01

    The document reproduces the text of the Project and Supply Agreement which was approved by the Agency's Board of Governors on 15 June 1990, and concluded on 20 September 1990, among the Agency and the Governments of the Republic of Zaire and the United States of America for the transfer of enriched uranium for a research reactor in Zaire. The supplied material will be an instrumented fuel element containing uranium enriched to 19.98 percent in the isotope uranium-235. 1 tab

  7. Uranium conversion; Urankonvertering

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

    2006-03-15

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF{sub 6} and UF{sub 4} are present require equipment that is made of corrosion resistant material.

  8. The opportunities for uranium development in South Australia

    International Nuclear Information System (INIS)

    Jackson, N.

    1979-07-01

    The opportunities for uranium development in South Australia are discussed. The author outlines the likely development of three known uranium deposits, shows the world energy and uranium requirements and makes some observations on uranium enrichment

  9. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  10. Uranium industry in the USSR

    International Nuclear Information System (INIS)

    Nikipelov, B.V.; Chernov, A.G.

    1990-01-01

    A brief historical account of the Soviet production of natural and enriched uranium is given. The geological and geographical location of major uranium deposits are mentioned. The processing of natural ores including in-situ leaching (ISL) is also briefly described. Gas centrifuges play a large part in uranium enrichment. The role of Techsnabexport for the export of nuclear materials is explained

  11. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Rasmussen, B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO 2 F 2 deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO 2 F 2 deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO 2 F 2 source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO 2 F 2 deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined

  12. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO{sub 2}F{sub 2} deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO{sub 2}F{sub 2} deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO{sub 2}F{sub 2} source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO{sub 2}F{sub 2} deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25

  13. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2009-01-01

    The Director General has received a letter dated 16 July 2009 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2008. 2. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2008

  14. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2013-01-01

    The Secretariat has received a note verbale dated 2 July 2013 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2012. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2012 [fr

  15. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2011-01-01

    The Director General has received a note verbale dated 29 April 2011 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2010. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2010 [fr

  16. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of Highly Enriched Uranium

    International Nuclear Information System (INIS)

    2007-01-01

    The Director General has received a Note Verbale dated 3 July 2007 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2006. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil highly enriched uranium (HEU) as of 31 December 2006 [fr

  17. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2012-01-01

    The Secretariat has received a note verbale dated 20 September 2012 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2011. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2011 [fr

  18. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2012-01-01

    The Director General has received a note verbale dated 14 October 2010 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2009. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2009 [es

  19. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of Highly Enriched Uranium

    International Nuclear Information System (INIS)

    2007-01-01

    The Director General has received a Note Verbale dated 3 July 2007 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2006. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil highly enriched uranium (HEU) as of 31 December 2006 [es

  20. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2012-01-01

    The Secretariat has received a note verbale dated 20 September 2012 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2011. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2011 [es

  1. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2011-01-01

    The Director General has received a note verbale dated 29 April 2011 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2010. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2010 [es

  2. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2013-01-01

    The Secretariat has received a note verbale dated 2 July 2013 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2012. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2012 [es

  3. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2009-01-01

    The Director General has received a letter dated 16 July 2009 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2008. 2. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2008 [es

  4. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2010-01-01

    The Director General has received a note verbale dated 14 October 2010 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2009. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2009

  5. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of Highly Enriched Uranium

    International Nuclear Information System (INIS)

    2007-01-01

    The Director General has received a Note Verbale dated 3 July 2007 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2006. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil highly enriched uranium (HEU) as of 31 December 2006

  6. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2012-01-01

    The Secretariat has received a note verbale dated 20 September 2012 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2011. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2011

  7. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    2013-01-01

    The Secretariat has received a note verbale dated 2 July 2013 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2012. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2012

  8. The ten stages pilot plant: its utilization in the research of the dynamic behavior and regullating process of a jet nozzle cascade for uranium enrichment

    International Nuclear Information System (INIS)

    Yadoya, R.; Camara, A.S.; Consiglio, R.V.; Bley, P.; Hein, H.; Linder, G.

    1986-01-01

    A ten stage pilot plant to study experimentally dynamic behavior of a uranium enrichment plant based on separation nozzle process was developed and constructed at Karlsruhe Nuclear Research Center. This installation was transfered to the Development Center of Nuclear Technology (CDTN) of Nuclebras in Belo Horizonte, Brazil. The separation elements installed have a new design with higher efficiency, Known as double-deflections system. The power plant has been used to improve the control method and to prove the stability of separation nozzle cascade under pertubations produced artificially. The stabilization process of UF 6 quantity in cascade by UF 6 inventory regulation at bottom stage will have practication in the First Cascade, in Rezende, RJ, Brazil and may be uded i emonstration plant. The experimental results have shown to be comparable with those obtained by computer simulation. (Author) [pt

  9. The text of the Agreement of 2 December 1983 concerning the Agency's assistance to Morocco for the transfer of enriched uranium from the United States for a research reactor

    International Nuclear Information System (INIS)

    1984-01-01

    The full text of the Project and Supply Agreement of 2 December 1983 between the Agency and the Government of the Kingdom of Morocco and the United States of America for the transfer of moderately enriched uranium for a research reactor in Morocco is presented

  10. Calculations of received dose for different points in the enrichment uranium oxide warehouse at 4%; Calculos de dosis recibida para diferentes puntos en el almacen de oxido de uranio enriquecido al 4%

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1990-06-15

    In order to verifying that the received dose so much inside as outside of the warehouse of enriched uranium dioxide to 4% it doesn't represent risk to the personnel, the modelling of this and the corresponding calculations for the extreme case of dose at contact are made. (Author)

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

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Fishbone, L.G.; Moussalli, G.; Naegele, G.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-06-01

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

  14. Uranium industry annual 1994

    International Nuclear Information System (INIS)

    1995-01-01

    The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data collected on the ''Uranium Industry Annual Survey'' (UIAS) provide a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ''Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,'' is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2

  15. Uranium industry annual 1998

    International Nuclear Information System (INIS)

    1999-01-01

    The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data provides a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ''Uranium Industry Annual Survey'' is provided in Appendix C. The Form EIA-858 ''Uranium Industry Annual Survey'' is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs

  16. Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

    2007-10-01

    The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department of Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor

  17. Validation of KENO V.a for criticality safety calculations of low-enriched uranium-235 systems

    International Nuclear Information System (INIS)

    McCamis, R.H.

    1991-02-01

    The criticality safety analysis program KENO V.a, together with a 27-energy-group ENDF/B-IV criticality safety cross-section library, has been validated by comparison of calculations with the experimental results from critical benchmarks dealing with low-enriched (≤ 5 wt%) 235 U systems, obtained both from the literature and from recent AECL Research experiments with the SLOWPOKE Demonstration Reactor. The combination of the code and this data library is shown to be very suitable for criticality safety analyses of low-enriched 235 U systems, with mean values of the calculated reactivities being within 1% of the experimental values. (6 figs., 3 tabs., 37 refs.)

  18. Analysis of organizational options for the uranium enrichment enterprise in relation to asset divesture. [BPA; TVA; SYNFUELS; CONRAIL; British TELECOM; COMSTAT

    Energy Technology Data Exchange (ETDEWEB)

    Harrer, B.J.; Hattrup, M.P.; Dase, J.E.; Nicholls, A.K.

    1986-08-01

    This report presents a comparison of the characteristics of some prominent examples of independent government corporations and agencies with respect to the Department of Energy's (DOE) uranium enrichment enterprise. The six examples studied were: the Bonneville Power Administration (BPA); the Tennessee Valley Authority (TVA); the Synthetic Fuels Corporation (SYNFUELS); the Consolidated Rail Corporation (CONRAIL); the British Telecommunications Corporation (British TELECOM); and the Communications Satellite Organization (COMSAT), in order of decreasing levels of government ownership and control. They range from BPA, which is organized as an agency within DOE, to COMSAT, which is privately owned and free from almost all regulations common to government agencies. Differences in the degree of government involvement in these corporations and in many other characteristics serve to illustrate that there are no accepted standards for defining the characteristics of government corporations. Thus, historical precedent indicates considerable flexibility would be available in the development of enabling legislation to reorganize the enrichment enterprise as a government corporation or independent government agency.

  19. Instrument calibration and measurement plan for the poorly measured/unmeasured category of highly enriched uranium at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Glosup, J.; Mount, M. E.

    1999-01-01

    In partial response to a Department of Energy (DOE) request to evaluate the state of measurements of special nuclear material, Lawrence Livermore National Laboratory (LLNL) evaluated and classified all highly enriched uranium (HEU) metal and oxide items in its inventory. Because of a lack of traceable HEU standards, no items were deemed to fit the category of well measured. A subsequent DOE-HQ sponsored survey by New Brunswick Laboratory resulted in their preparation of a set of certified reference material (CRM) standards for HEU oxide (U 3 O 8 ) that are projected for delivery during September of 1999. However, CRM standards for HEU metal are neither in preparation nor are they expected to be prepared within the foreseeable future. Consequently, HEU metal working standards must be developed if the poorly measured/unmeasured portion of the LLNL inventory is to be reclassified. This paper describes the approach that LLNL will take to (1) develop a set of HEU metal working standards; (2) develop HEU metal and oxide calibration curves for the passive-active neutron (PAN) shuffler that are functions of mass, enrichment, size, and shape; and (3) reclassify the poorly measured/unmeasured inventory through direct measurement or reprocessing of previously archived data

  20. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  1. Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

    Directory of Open Access Journals (Sweden)

    Chihyung Kim

    2016-04-01

    Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

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

    International Nuclear Information System (INIS)

    Nazare, S.

    1982-07-01

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U 3 Si and U 3 Si 2 ) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U 3 Si- und U 3 Si-Al up to U-densities of 6.0 g U/cm 3 . The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.) [de

  3. An Assessment of the Detection of Highly Enriched Uranium and its Use in an Improvised Nuclear Device using the Monte Carlo Computer Code MCNP-5

    Science.gov (United States)

    Cochran, Thomas

    2007-04-01

    In 2002 and again in 2003, an investigative journalist unit at ABC News transported a 6.8 kilogram metallic slug of depleted uranium (DU) via shipping container from Istanbul, Turkey to Brooklyn, NY and from Jakarta, Indonesia to Long Beach, CA. Targeted inspection of these shipping containers by Department of Homeland Security (DHS) personnel, included the use of gamma-ray imaging, portal monitors and hand-held radiation detectors, did not uncover the hidden DU. Monte Carlo analysis of the gamma-ray intensity and spectrum of a DU slug and one consisting of highly-enriched uranium (HEU) showed that DU was a proper surrogate for testing the ability of DHS to detect the illicit transport of HEU. Our analysis using MCNP-5 illustrated the ease of fully shielding an HEU sample to avoid detection. The assembly of an Improvised Nuclear Device (IND) -- a crude atomic bomb -- from sub-critical pieces of HEU metal was then examined via Monte Carlo criticality calculations. Nuclear explosive yields of such an IND as a function of the speed of assembly of the sub-critical HEU components were derived. A comparison was made between the more rapid assembly of sub-critical pieces of HEU in the ``Little Boy'' (Hiroshima) weapon's gun barrel and gravity assembly (i.e., dropping one sub-critical piece of HEU on another from a specified height). Based on the difficulty of detection of HEU and the straightforward construction of an IND utilizing HEU, current U.S. government policy must be modified to more urgently prioritize elimination of and securing the global inventories of HEU.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  6. Uranium recovery from slags of metallic uranium

    International Nuclear Information System (INIS)

    Fornarolo, F.; Frajndlich, E.U.C.; Durazzo, M.

    2006-01-01

    The Center of the Nuclear Fuel of the Institute of Nuclear Energy Research - IPEN finished the program of attainment of fuel development for research reactors the base of Uranium Scilicet (U 3 Si 2 ) from Hexafluoride of Uranium (UF 6 ) with enrichment 20% in weight of 235 U. In the process of attainment of the league of U 3 Si 2 we have as Uranium intermediate product the metallic one whose attainment generates a slag contend Uranium. The present work shows the results gotten in the process of recovery of Uranium in slags of calcined slags of Uranium metallic. Uranium the metallic one is unstable, pyrophoricity and extremely reactive, whereas the U 3 O 8 is a steady oxide of low chemical reactivity, what it justifies the process of calcination of slags of Uranium metallic. The calcination of the Uranium slag of the metallic one in oxygen presence reduces Uranium metallic the U 3 O 8 . Experiments had been developed varying it of acid for Uranium control and excess, nitric molar concentration gram with regard to the stoichiometric leaching reaction of temperature of the leaching process. The 96,0% income proves the viability of the recovery process of slags of Uranium metallic, adopting it previous calcination of these slags in nitric way with low acid concentration and low temperature of leaching. (author)

  7. Radioactive ground-water contamination from an enriched-uranium cold scrap recovery operation, Wood River Junction, Rhode Island

    International Nuclear Information System (INIS)

    Ryan, B.J.; Kipp, K.L. Jr.

    1984-01-01

    Liquid wastes from a uranium-bearing cold scrap recovery plant at an industrial site in Wood River Junction, Rhode Island were discharged to the environment through evaporation ponds from 1966 to 1980. Leakage from the polyethylene- and polyvinylchloride-lined ponds resulted in a plume of contaminated ground water that extends from the ponds northwestward to the Pawcatuck River through a highly permeable sand and gravel aquifer of glacial origin. Contaminants include: strontium 90, technetium 99, boron, nitrate and potassium. Water quality data from more than 100 observation wells indicate that the plume of contamination is approximately 700 meters long, 100 meters wide, and is confined to the upper 25 meters of saturated thickness where sediments consist of medium to coarse sand and gravel. No contamination has been detected in fine sands and silts underlying the coarser materials. Piezometric-head and water-quality data from wells screened at multiple depths on both sides of the river indicate that contaminants discharge both to the river and to a swampy area at the west edge of the river. Dilution precludes detection of contaminants once they have entered the river, which has an average flow of 5 cubic meters per second

  8. Nuclear facilities of the National Academy of Sciences of Belarus on the basis of highly enriched uranium

    International Nuclear Information System (INIS)

    Chigrinov, S.; Bournos, V.; Serafimovich, I.; Fokov, Yu.; Routkovskaia, C.; Voropay, N.; Kiyavitskaya, H.

    2005-01-01

    The investigations in the field of nuclear physics, development of numerical calculation methods for nuclear reactors, neutron physics and etc. are carried out at the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) since the 60s after putting into operation the research reactor and the critical assemblies. A large range of different configuration (geometry,composition) of critical assemblies have been constructed at the NAS Belarus during 25 years of studying neutronic of the special (fast and thermal) reactors. The Chernobyl accident brought a massive public reaction to nuclear efforts and the reactor ceased operation in 1987 and was shut down in 1991 and at present all investigations in these fields are being carried out on the basis of the subcriticall assemblies driven with high intensity neutron generator. The facilities with fast and thermal neutron spectra are fuelled with UO 2 and Umet. enriched to 10% -90% in 235 U. (author)

  9. Verification experiment on the downblending of high enriched uranium (HEU) at the Portsmouth Gaseous Diffusion Plant. Digital video surveillance of the HEU feed stations

    International Nuclear Information System (INIS)

    Martinez, R.L.; Tolk, K.; Whiting, N.; Castleberry, K.; Lenarduzzi, R.

    1998-01-01

    As part of a Safeguards Agreement between the US and the International Atomic Energy Agency (IAEA), the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, was added to the list of facilities eligible for the application of IAEA safeguards. Currently, the facility is in the process of downblending excess inventory of HEU to low enriched uranium (LEU) from US defense related programs for commercial use. An agreement was reached between the US and the IAEA that would allow the IAEA to conduct an independent verification experiment at the Portsmouth facility, resulting in the confirmation that the HEU was in fact downblended. The experiment provided an opportunity for the DOE laboratories to recommend solutions/measures for new IAEA safeguards applications. One of the measures recommended by Sandia National Laboratories (SNL), and selected by the IAEA, was a digital video surveillance system for monitoring activity at the HEU feed stations. This paper describes the SNL implementation of the digital video system and its integration with the Load Cell Based Weighing System (LCBWS) from Oak Ridge National Laboratory (ORNL). The implementation was based on commercially available technology that also satisfied IAEA criteria for tamper protection and data authentication. The core of the Portsmouth digital video surveillance system was based on two Digital Camera Modules (DMC-14) from Neumann Consultants, Germany

  10. Proposed rulemaking exemption from licensing requirements for smelted alloys containing residual technetium-99 and low-enriched uranium. Draft environmental statement

    International Nuclear Information System (INIS)

    1980-10-01

    This environmental statement is concerned with the costs and benefits to be incurred in the exemption from federal (NRC) licensing requirements for smelted alloys containing residual technetium-99 and low enriched uranium at concentrations of 5 ppM and 17.5 ppM, respectively. The contaminated metals considered explicitly in the statement for smelting are the scrap copper, nickel, and iron resulting from the improvement and upgrading of the Department of Energy's gaseous diffusion plants. The environmental impacts considered are primarily the radiological doses derived from the smelted metal by the general public including foundary and fabrication workers. Economic impacts associated with building and operating the facilities to smelt the scrap are also considered. In addition, the energy savings through recycling of these materials back into the commercial marketplace instead of burying them is estimated. The principal alternatives to this proposed action are the sale of surface decontaminated scrap, recycle of decontaminated equipment, burial of contaminated scrap, surface storage of contaminated scrap; and issuance of either a general or specific license for scrap disposal

  11. Communication received from France concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of highly enriched uranium

    International Nuclear Information System (INIS)

    2003-01-01

    The Director General has received a Note Verbale, dated 2 September 2003, from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2002. The Government of France has also made available a statement of its annual figures for holdings of civil high-enriched uranium (HEU) as of 31 December 2002. In light of the request expressed by the Government of France in its Note Verbale of 28 November 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the enclosures of the Note Verbale of 2 September 2003 are attached for the information of all Member States

  12. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  13. Acceptable standard format and content for the fundamental nuclear material control (FNMC) plan required for low-enriched uranium facilities. Revision 2

    International Nuclear Information System (INIS)

    Joy, D.R.

    1995-12-01

    This report documents a standard format suggested by the NRC for use in preparing fundamental nuclear material control (FNMC) plans as required by the Low Enriched Uranium Reform Amendments (10CFR 74.31). This report also describes the necessary contents of a comprehensive plan and provides example acceptance criteria which are intended to communicate acceptable means of achieving the performance capabilities of the Reform Amendments. By using the suggested format, the licensee or applicant will minimize administrative problems associated with the submittal, review and approval of the FNMC plan. Preparation of the plan in accordance with this format Will assist the NRC in evaluating the plan and in standardizing the review and licensing process. However, conformance with this guidance is not required by the NRC. A license applicant who employs a format that provides a equal level of completeness and detail may use their own format. This document is also intended for providing guidance to licensees when making revisions to their FNMC plan

  14. Communication received from Germany concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of high enriched uranium

    International Nuclear Information System (INIS)

    2005-01-01

    The Director General has received a letter dated 18 April 2005 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2004. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of high enriched uranium (HEU) as of 31 December 2004. In light of the request expressed by the Federal Republic of Germany in its Note Verbale of 1 December 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the enclosures of the letter of 18 April 2005 are attached for the information of all Member States

  15. Communication received from France concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of high enriched uranium

    International Nuclear Information System (INIS)

    2005-01-01

    The Director General has received a Note Verbale dated 27 September 2005 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2003. The Government of France has also made available a statement of its annual figures for holdings of civil high-enriched uranium (HEU) as of 31 December 2004. In light of the request expressed by the Government of France in its Note Verbale of 28 November 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the enclosures of the Note Verbale of 27 September 2005 are attached for the information of all Member States

  16. Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

    International Nuclear Information System (INIS)

    Pierre, J.R.M.

    1996-01-01

    Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

  17. Communication dated 13 March 2009 received from the Resident Representative of the Russian Federation to the IAEA on the Russian initiative to establish a guaranteed reserve of low enriched uranium

    International Nuclear Information System (INIS)

    2009-01-01

    The Secretariat has received a communication dated 13 March 2009 from the Resident Representative of the Russian Federation, forwarding a statement delivered on 5 March 2009 by the Governor for the Russian Federation on the Board of Governors, Ambassador G.V. Berdennikov concerning the Russian initiative to establish a guaranteed reserve of low enriched uranium. As requested in that communication, the statement is circulated herewith for the information of Member States

  18. Project and supply agreement. The text of the agreement of 17 June 1993 between the International Atomic Energy Agency and the governments of Colombia and the United States of America concerning the transfer of enriched uranium for a research reactor

    International Nuclear Information System (INIS)

    1994-12-01

    The text of the Project and Supply Agreement, which was approved by the Agency's Board of Governors on 25 February 1993 and 2 December 1993, between the Agency and the Governments of Colombia and the United States of America for the transfer of enriched uranium for a research reactor is reproduced herein for the information of all Members. The agreement entered into force on 17 June 1994, pursuant to Article XII

  19. Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants

    Energy Technology Data Exchange (ETDEWEB)

    Robert Bean; Casey Durst

    2009-10-01

    This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a

  20. Pine Creek uranium province

    International Nuclear Information System (INIS)

    Bower, M.B.; Needham, R.S.; Page, R.W.; Stuart-Smith, P.G.; Wyborn, L.A.I.

    1985-01-01

    The objective of this project is to help establish a sound geological framework of the Pine Creek region through regional geological, geochemical and geophysical studies. Uranium ore at the Coronation Hill U-Au mine is confined to a wedge of conglomerate in faulted contact with altered volcanics. The uranium, which is classified as epigenetic sandstone type, is derived from a uranium-enriched felsic volcanic source