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1

Uranium Enrichment  

NLE Websites -- All DOE Office Websites (Extended Search)

Enrichment Depleted Uranium line line Uranium Enrichment Depleted Uranium Health Effects Uranium Enrichment A description of the uranium enrichment process, including gaseous...

2

NUCLEAR BOMBS FROM LOW- ENRICHED URANIUM OR “SPENT ” FUEL  

E-Print Network (OSTI)

Conventional wisdom says that low-enriched uranium is not suitable for making nuclear weapons. However, an article in USA Today claims that “rogue ” states and terrorists have discovered that this is untrue. Not only that, but terrorists could separate plutonium from irradiated fuel (often called “spent fuel”) more easily than previously thought. (584.5495) WISE Amsterdam – Lowenriched uranium (LEU) is uranium containing up to 20 % uranium-235. Uranium with higher enrichment levels is classified as high-enriched, and is subject to international safeguards because it can be used to make nuclear weapons. However, a USA Today article claims that “rogue countries and terrorists” have discovered that it is possible to make nuclear weapons with uranium of lower enrichment, according to classified nuclear threat reports (1). The information is not entirely new. Back in 1996, a standard book on nuclear weapons material stated, “a self-sustaining chain reaction in a nuclear weapon cannot occur in depleted or natural or low-enriched uranium and is only theoretically IN THIS ISSUE: possible in LEU of roughly 10 percent or greater ” (2). Fuel for nuclear power reactors would not be suitable – this is typically enriched to 3-5 % uranium-235. However, for a “rogue state” wanting to make high-enriched uranium, it would take less work to start with nuclear fuel than with natural uranium. It could be done in a “small and easy to hide ” uranium enrichment plant – perhaps similar to the plant which has recently been discovered in Iran (3). Nevertheless, it would still require a substantial operation, since the fuel would need to be converted to uranium hexafluoride, enriched and then reconverted to uranium metal. More significantly, many research reactors use uranium of just under

unknown authors

2003-01-01T23:59:59.000Z

3

The U.S. relies on foreign uranium, enrichment services to fuel ...  

U.S. Energy Information Administration (EIA)

The U.S. relies on foreign uranium, enrichment services to fuel its nuclear power plants. Source: U.S. Energy Information Administration, Uranium Marketing Annual Report.

4

Uranium Mining and Enrichment  

NLE Websites -- All DOE Office Websites (Extended Search)

Overview Presentation » Uranium Mining and Enrichment Overview Presentation » Uranium Mining and Enrichment Uranium Mining and Enrichment Uranium is a radioactive element that occurs naturally in the earth's surface. Uranium is used as a fuel for nuclear reactors. Uranium-bearing ores are mined, and the uranium is processed to make reactor fuel. In nature, uranium atoms exist in several forms called isotopes - primarily uranium-238, or U-238, and uranium-235, or U-235. In a typical sample of natural uranium, most of the mass (99.3%) would consist of atoms of U-238, and a very small portion of the total mass (0.7%) would consist of atoms of U-235. Uranium Isotopes Isotopes of Uranium Using uranium as a fuel in the types of nuclear reactors common in the United States requires that the uranium be enriched so that the percentage of U-235 is increased, typically to 3 to 5%.

5

Remote Inspection Devices for Spent Reactor Enriched Uranium Fuel Elements  

SciTech Connect

A remote video inspection was developed and deployed in Argentina for the detailed inspection of highly radioactive spent reactor fuel (SNF) as a prerequisite to its shipment to the Savannah River Site (SRS) in the United States for long-term storage and disposition. The fuel is highly enriched uranium (HEU) spent assemblies dating from 1967 to 1989 and aluminum clad uranium-aluminum alloy of a typical material test reactor design. The specialized video system was designed for low cost, high portability, easy setup, and ease of usage, while accommodating the differing electrical systems (i.e. 110/60 Hz, 220/50 Hz) between the United States and Argentina.

Heckendorn, F.M.

2001-01-03T23:59:59.000Z

6

Irradiation performance of low-enriched uranium fuel elements  

SciTech Connect

The status of the testing and evaluation of full-sized experimental low- and medium-enriched uranium fuel elements in the Oak Ridge Research Reactor is presented. Medium-enriched elements containing oxide and aluminide have been completely evaluated at burnups up to 75%. A low-enriched U/sub 3/Si/sub 2/ element has been evaluated at 41% burnup. Other silicide and oxide elements have completed irradiation satisfactorily to burnups of 75% and are now being evaluated. All results to date confirm the expected good performance of these elements in the medium power research reactor environment.

Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

1984-10-14T23:59:59.000Z

7

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

SciTech Connect

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.

Heckendorn, F.M.

2001-01-03T23:59:59.000Z

8

High uranium density dispersion fuel for the reduced enrichment of research and test reactors program.  

E-Print Network (OSTI)

??This work describes the fabrication of a high uranium density fuel for the Reduced Enrichment of Research and Test Reactors Program. In an effort to… (more)

[No author

2006-01-01T23:59:59.000Z

9

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

Science Conference Proceedings (OSTI)

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

Pesic, Milan P

2003-10-15T23:59:59.000Z

10

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

SciTech Connect

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

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

1997-03-28T23:59:59.000Z

11

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel  

NLE Websites -- All DOE Office Websites (Extended Search)

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Production: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel ... Fact Sheet Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel

12

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel  

National Nuclear Security Administration (NNSA)

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Production: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel ... Fact Sheet Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel

13

Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, 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 high-enriched uranium fuel core. These studies 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.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2009-11-01T23:59:59.000Z

14

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

Science Conference Proceedings (OSTI)

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

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

2010-02-01T23:59:59.000Z

15

Derived enriched uranium market  

SciTech Connect

The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market.

Rutkowski, E.

1996-12-01T23:59:59.000Z

16

Engineering analysis of low enriched uranium fuel using improved zirconium hydride cross sections  

E-Print Network (OSTI)

A neutronic and thermal hydraulic analysis of the 1-MW TRIGA research reactor at the Texas A&M University Nuclear Science Center using a new low enriched uranium fuel (named 30/20 fuel) was completed. This analysis provides safety assessment for the change out of the existing high enriched uranium fuel to this high-burnup, low enriched uranium fuel design. The codes MCNP and Monteburns were utilized for the neutronic analysis while the code PARET was used to determine fuel and cladding temperatures. All of these simulations used improved zirconium hydride cross sections that were provided by Dr. Ayman Hawari at North Carolina State University. The neutronic and thermal analysis showed that the reactor will operate with approximately the same fuel lifetime as the current high enriched uranium fuel and stay within the thermal and safety limits for the facility. It was also determined that the control rod worths and the temperature coefficient of reactivity would provide sufficient negative reactivity to control the reactor during the fuel�s complete lifetime. An assessment of the fuel�s viability for use with the Advanced Fuel Cycle Initiative�s Reactor Accelerator Coupling Experiments program was also performed. The objective of this study was to confirm the continued viability of these experiments with the reactor operating using this new fuel. For these experiments, the accelerator driven system must produce fission heating in excess of 1 kW when driven by a 20 kW accelerator system. This criterion was met using the new fuel. Therefore the change out of the fuel will not affect the viability of these experiments.

Candalino, Robert Wilcox

2006-08-01T23:59:59.000Z

17

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

SciTech Connect

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.

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

2010-10-01T23:59:59.000Z

18

PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP  

Science Conference Proceedings (OSTI)

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01T23:59:59.000Z

19

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

SciTech Connect

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.

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

1986-01-01T23:59:59.000Z

20

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

Science Conference Proceedings (OSTI)

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 establishing preconceptual fabrication facility designs.

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

2008-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

SciTech Connect

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

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

1985-01-01T23:59:59.000Z

22

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania  

SciTech Connect

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

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

2010-07-01T23:59:59.000Z

23

Reference (Axially Graded) Low Enriched Uranium Fuel Design for the High Flux Isotope Reactor (HFIR)  

Science Conference Proceedings (OSTI)

During the past five years, staff at the Oak Ridge National Laboratory (ORNL) have studied the issue of whether the HFIR could be converted to low enriched uranium (LEU) fuel without degrading the performance of the reactor. Using state-of-the-art reactor physics methods and behind-the-state-of-the-art thermal hydraulics methods, the staff have developed fuel plate designs (HFIR uses two types of fuel plates) that are believed to meet physics and thermal hydraulic criteria provided the reactor power is increased from 85 to 100 MW. The paper will present a defense of the results by explaining the design and validation process. A discussion of the requirements for showing applicability of analyses to approval for loading the fuel to HFIR lead test core irradiation currently scheduled for 2016 will be provided. Finally, the potential benefits of upgrading thermal hydraulics methods will be discussed.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

24

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

SciTech Connect

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.

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

2009-07-01T23:59:59.000Z

25

Overview: A Legacy of Uranium Enrichment  

NLE Websites -- All DOE Office Websites (Extended Search)

A Legacy of Uranium Enrichment Depleted Uranium is a Legacy of Uranium Enrichment Cylinders Photo Next Screen Management Responsibilities...

26

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

Science Conference Proceedings (OSTI)

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.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01T23:59:59.000Z

27

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

SciTech Connect

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

Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

2010-07-01T23:59:59.000Z

28

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

SciTech Connect

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.

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

2012-03-01T23:59:59.000Z

29

CRITICAL EXPERIMENTS ON SLIGHTLY ENRICHED URANIUM METAL FUEL ELEMENTS IN GRAPHITE LATTICES  

SciTech Connect

A series of clean critical experiments was performed in the SGR critical facility utilizing 2 wt% enriched, uranium metal, hollow cylinder, fuel elements in AGOT graphite moderator. Six lattice spacings were used, varying from 6.93 to 16.0 in. on a triangular pitch. Critical loadings and fuel element worths were determined and compared to the results of 4-group diffusion theory. Calculations utilized TEMPEST, S/sub 4/, FORM, and AIM-5 programs on the IBM 7090. The calculated K/sub eff/ compared well with experiments over the full range of moderator-to-fuel volume ratios when using a 2200 m/sec graphite absorption cross section of 4.07 mb. The sensitivity of the calculation to variations in the graphite absorption cross section was examined and the experimental error due to inventory uncertainties was assessed. The differential worths of both the central and peripheral fuel elements were obtained and agreed in general with AIM- 5 calculations. The thermal flux traverse of a unit cell was shown to agree best with a Wilkins' spectrum option of TEMPEST. Details of both the experimental and theoretical methods are given. (auth) The work functions of cesiated and cesium- hydridecoated surfaces are studied. A thermionlc cell for performance analyses is described. Design characteristics of water-cooled and liquid-metal-cooled nuclearthermionic generators for naval power applications are compared. (T.F.H.)

Campbell, R.W.; Doyas, R.J.; Field, H.C.; Guderjahn, C.A.; Guenther, R.L.; Hausknecht, D.E.; Mayer, M.S.; Morewitz, H.A.

1963-06-30T23:59:59.000Z

30

Highly Enriched Uranium Transparency Program  

NLE Websites -- All DOE Office Websites (Extended Search)

and Climate Research Center for Geospatial Analysis Program Highlights Index Highly Enriched Uranium Transparency Program EVS staff members helped to implement transparency and...

31

Production and Handling Slide 37: The Uranium Fuel Cycle  

NLE Websites -- All DOE Office Websites (Extended Search)

Table of Contents The Uranium Fuel Cycle Refer to caption below for image description The enrichment process generates two streams of uranium hexafluoride, one enriched in...

32

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

Science Conference Proceedings (OSTI)

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 supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

SCHWINKENDORF, K.N.

2006-05-12T23:59:59.000Z

33

Production and Handling Slide 43: The Uranium Fuel Cycle  

NLE Websites -- All DOE Office Websites (Extended Search)

Presentation Table of Contents The Uranium Fuel Cycle Refer to caption below for image description Enriched uranium hexafluoride, generally containing 3 to 5% uranium-235, is sent...

34

Uranium Enrichment Decontamination and Decommissioning Fund's...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit, OAS-FS-10-05 Uranium Enrichment Decontamination and...

35

Assumptions and criteria for performing a feasibility study of the conversion of the high flux isotope reactor core to use low-enriched uranium fuel  

SciTech Connect

This paper provides a preliminary estimate of the operating power for the High Flux Isotope Reactor when fuelled with low enriched uranium (LEU). Uncertainties in the fuel fabrication and inspection processes are reviewed for the current fuel cycle [highly enriched uranium (HEU)] and the impact of these uncertainties on the proposed LEU fuel cycle operating power is discussed. These studies indicate that for the power distribution presented in a companion paper in these proceedings, the operating power for an LEU cycle would be close to the current operating power. (authors)

Primm Iii, R. T. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6399 (United States); Ellis, R. J.; Gehin, J. C. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6172 (United States); Moses, D. L. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6050 (United States); Binder, J. L. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6162 (United States); Xoubi, N. [Univ. of Cincinnati, Rhodes Hall, ML 72, PO Box 210072, Cincinnati, OH 45221-0072 (United States)

2006-07-01T23:59:59.000Z

36

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

Science Conference Proceedings (OSTI)

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 bounds of known technology and are adaptable to the high-volume production required to process {approx} 2.5 to 4 tons of U/Mo and produce {approx}16,000 flat plates for U.S. reactors annually ({approx}10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M.

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

2007-09-30T23:59:59.000Z

37

U. S. forms uranium enrichment corporation  

SciTech Connect

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

Seltzer, R.

1993-07-12T23:59:59.000Z

38

Highly Enriched Uranium Materials Facility | Y-12 National Security...  

NLE Websites -- All DOE Office Websites (Extended Search)

Highly Enriched Uranium ... Highly Enriched Uranium Materials Facility HEUMF The Highly Enriched Uranium Materials Facility is our nation's central repository for highly enriched...

39

Development of a low enrichment uranium core for the MIT reactor  

E-Print Network (OSTI)

An investigation has been made into converting the MIT research reactor from using high enrichment uranium (HEU) to low enrichment uranium (LEU) with a newly developed fuel material. The LEU fuel introduces negative ...

Newton, Thomas Henderson

2006-01-01T23:59:59.000Z

40

EA-1172: Sale of Surplus Natural and Low Enriched Uranium, Piketon, Ohio  

Energy.gov (U.S. Department of Energy (DOE))

This EA evaluates the environmental impacts for the proposal to sell uranium for subsequent enrichment and fabrication into commercial nuclear power reactor fuel.  The uranium is currently stored...

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Profile of World Uranium Enrichment Programs-2009  

Science Conference Proceedings (OSTI)

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 demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency and the European Atomic Energy Community began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access inspections in cascade halls, and the use of nondestructive assay measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. Uranium enrichment capacity has continued to expand on all fronts in the last few years. GCEP capacity is expanding in anticipation of the eventual shutdown of the less-efficient GDPs, the termination of the U.S.-Russia HEU blend-down program slated for 2013, and the possible resurgence of nuclear reactor construction as part of an expected 'Nuclear Renaissance'. Overall, a clear trend in the world profile of uranium enrichment plant operation is the continued movement towards multinational projects driven by commercial and economic interests. Along this vein, the safeguards community is continuing to develop new safeguards techniques and technologies that are not overly burdensome to enrichment plant operators while delivering more effective and efficient results. This report provides a snapshot overview of world enrichment capacity in 2009, including profiles of the uranium enrichment programs of individual states. It is a revision of a 2007 report on the same topic; significant changes in world enrichment programs between the previous and current reports are emphasized. It is based entirely on open-source information, which is dependent on published sources and may theref

Laughter, Mark D [ORNL

2009-04-01T23:59:59.000Z

42

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

E-Print Network (OSTI)

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. ...

American Society for Testing and Materials. Philadelphia

2000-01-01T23:59:59.000Z

43

Uncertainty clouds uranium enrichment corporation's plans  

SciTech Connect

An expected windfall to the US Treasury from the sale of the Energy Dept.'s commercial fuel enrichment facilities may evaporate in the next few weeks when the Clinton administration submits its fiscal 1994 budget proposal to Congress, according to congressional and administration officials. Under the Energy Policy Act of 1992, DOE is required to lease two uranium enrichment facilities, Portsmouth, Ohio, and Paducah, KY., to the government-owned US Enrichment Corp. (USEC) by July 1. Estimates by OMB and Treasury indicate a potential yearly payoff of $300 million from the government-owned company's sale of fuel for commercial reactors. Those two facilities use a process of gaseous diffusion to enrich uranium to about 3 percent for use as fuel in commercial power plants. DOE has contracts through at least 1996 to provide about 12 million separative work units (SWUs) yearly to US utilities and others world-wide. But under an agreement signed between the US and Russia last August, at least 10 metric tons, or 1.5 million SWUs, of low-enriched uranium (LEU) blended down from Russia warheads is expected to be delivered to the US starting in 1994. It could be sold at $50 to $60 per SWU, far below what DOE currently charges for its SWUs - $135 per SWU for 70 percent of the contract price and $90 per SWU for the remaining 30 percent.

Lane, E.

1993-03-24T23:59:59.000Z

44

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

SciTech Connect

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

Travelli, A.

1983-01-01T23:59:59.000Z

45

Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thicknesses of 1.27 or 1.52 mm. Fuel core dimensions varied according to the overall plate thicknesses with a minimum clad thickness of 0.20 mm. Tested fuels included UAl/sub x/, UAl/sub 2/, U/sub 3/O/sub 8/, U/sub 3/SiAl, U/sub 3/Si, U/sub 3/Si/sub 1.5/, U/sub 3/Si/sub 2/, U/sub 3/SiCu, USi, U/sub 6/Fe, and U/sub 6/Mn/sub 1.3/ materials. Although most miniplates were made with low-enriched uranium (19.9%), some with medium-enriched uranium (40 to 45%), a few with high-enriched uranium (93%), and a few with depleted uranium (0.2 to 0.4%) were tested for comparison. These fuel materials were irradiated to burnups ranging from /approximately/27 to 98 at. % /sup 235/U depletion. Operation of the experiment, measurement of miniplate thickness as the irradiation progressed, ultimate shipment of the irradiated miniplates to various hot cells, and preliminary results are reported here. 18 refs., 12 figs., 7 tabs.

Senn, R.L.

1989-04-01T23:59:59.000Z

46

Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center  

Science Conference Proceedings (OSTI)

The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

Myers, Astasia [Stanford University, Stanford, CA 94305, USA and MonAme Scientific Research Center, Ulaanbaatar (Mongolia)

2011-06-28T23:59:59.000Z

47

THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE  

E-Print Network (OSTI)

Soubbaramayer, (1979) in "Uranium Enrichment", S. Villani,and Davies, E. (1973) "Uranium Enrichment by Gas Centrifuge"THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

Olander, Donald R.

2013-01-01T23:59:59.000Z

48

THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE  

E-Print Network (OSTI)

Soubbaramayer, (1979) in "Uranium Enrichment", S. Villani,and Davies, E. (1973) "Uranium Enrichment by Gas Centrifuge"Nuclear Energy THE THEORY OF URANIUM ENRICHMENT BY THE GAS

Olander, Donald R.

2013-01-01T23:59:59.000Z

49

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

SciTech Connect

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.

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

2006-02-01T23:59:59.000Z

50

Profile of World Uranium Enrichment Programs - 2007  

SciTech Connect

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 of the future, but has yet to be demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency (IAEA) and the European Atomic Energy Community (EURATOM) began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access (LFUA) inspections in cascade halls, and the use of nondestructive assay (NDA) measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. This report provides a snapshot overview of world enrichment capacity in 2007, including profiles of the uranium enrichment programs of individual states. It is based on open-source information, which is dependent on unclassified sources and may therefore not reflect the most recent developments. In addition, it briefly describes some of the safeguards techniques being used at various enrichment plants, including implementation of HSP recommendations.

Laughter, Mark D [ORNL

2007-11-01T23:59:59.000Z

51

Production and Handling Slide 23: The Uranium Fuel Cycle  

NLE Websites -- All DOE Office Websites (Extended Search)

Presentation Table of Contents The Uranium Fuel Cycle Refer to caption below for image description The fourth major step in the uranium fuel cycle is uranium enrichment. Slide 23...

52

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

SciTech Connect

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.

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

2008-07-01T23:59:59.000Z

53

Highly Enriched Uranium Transparency Program | National Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

Highly Enriched Uranium Transparency Program | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy...

54

The uranium cylinder assay system for enrichment plant safeguards  

Science Conference Proceedings (OSTI)

Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF{sub 6} cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF{sub 6} cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

Miller, Karen A [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Marlow, Johnna B [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory; Iwamoto, Tomonori [JNFL; Tamura, Takayuki [JNFL; Aiuchi, Syun [JNFL

2010-01-01T23:59:59.000Z

56

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

SciTech Connect

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

Conway, K. C.

2002-02-28T23:59:59.000Z

57

Cross section generation and physics modeling in a feasibility study of the conversion of the high flux isotope reactor core to use low-enriched uranium fuel  

SciTech Connect

A computational study has been initiated at ORNL to examine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The current study is limited to steady-state, nominal operation and are focused on the determination of the fuel requirements, primarily density, that are required to maintain the performance of the reactor. Reactor physics analyses are reported for a uranium-molybdenum alloy that would be substituted for the current fuel - U{sub 3}O{sub 8} mixed with aluminum. An LEU core design has been obtained and requires an increase in {sup 235}U loading of a factor of 1.9 over the current HEU fuel. These initial results indicate that the conversion from HEU to LEU results in a reduction of the thermal fluxes in the central flux trap region of approximately 9 % and in the outer beryllium reflector region of approximately 15%. Ongoing work is being performed to improve upon this initial design to further minimize the impact of conversion to LEU fuel. (authors)

Ellis, R. J.; Gehin, J. C.; Primm Iii, R. T. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2006-07-01T23:59:59.000Z

58

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

SciTech Connect

This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF{sub 6} and a (2) blend the pure HEU UF{sub 6} with diluent UF{sub 6} to produce LWR grade LEU-UF{sub 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.

1995-07-05T23:59:59.000Z

59

Low-Enrichment Fuel Development Program  

SciTech Connect

The national program of the Department of Energy at Argonne National Laboratory for the development of highly loaded uranium fuels, which provide the means for enrichment reduction, has been briefly described. The objectives of > 60 wt % uranium in plate-type fuels and greater than or equal to 45 wt % uranium in U--ZrH/sub x/ rod-type fuels are expected to be met. The most promising fuels will be further evaluated in full-size element irradiations and whole-core demonstrations on the route toward commercialization.

Stahl, D.

1978-01-01T23:59:59.000Z

60

Uranium enrichment in the United States  

SciTech Connect

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)

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

1975-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties  

E-Print Network (OSTI)

The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

Chiang, Keng-Yen

2012-01-01T23:59:59.000Z

62

Overview of reduced-enrichment fuels - development  

SciTech Connect

The US Reduced Enrichment Research and Test Reactor (RERTR) Program was established in 1978 to provide the technical means to operate research and test reactors with low-enrichment uranium (LEU) fuels without significant penalty in experiment performance, operation costs, component modifications, or safety characteristics. A large increase in /sup 238/U is required to reduce the enrichment, and a 10 to 15% increase in /sup 235/U is required to compensate for the extra absorption in /sup 238/U. The additional uranium can be accommodated by redesigning the fuel element to increase the fuel volume fraction in the reactor core and/or by increasing the uranium density in the fuel meat. Since fuel element redesign coupled with the highest density fuel available in 1978 is sufficient for only a few reactors, a fuel development and testing effort was begun to qualify much higher density fuels. The greatest emphasis has been on plate-type fuels, since plate-type reactors are the largest users of highly enriched uranium (HEU). In addition to the RERTR program's work with plate-type dispersion fuels, the CEA developed and tested the caramel fuel, consisting of sintered UO/sub 2/ wafers in Zircaloy-clad plates; GA Technologies developed highly loaded UZrH/sub x/ fuel for TRIGA reactors and tested it in cooperation with the RERTR Program; and Atomic Energy of Canada Ltd. developed and tested rod-type uranium silicide-Al dispersion fuel. The dispersion fuels were irradiated to high burnups to establish their limits of usability. A whole-core demonstration has been conducted in the ORR using 4.8 Mg U/m/sup 3/ U/sub 3/Si/sub 2/ dispersion fuel. Twenty-nine elements have achieved average burnups in excess of 40%.

Snelgrove, J.L.

1987-01-01T23:59:59.000Z

63

Surplus Highly Enriched Uranium Disposition Program plan  

SciTech Connect

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.

1996-10-01T23:59:59.000Z

64

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

SciTech Connect

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade 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. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

1995-07-05T23:59:59.000Z

65

SALE OF ENRICHED URANIUM AT THE FERNALD ENVIRONMENTAL MANAGEMENT...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Power Marketing Administration Other Agencies You are here Home SALE OF ENRICHED URANIUM AT THE FERNALD ENVIRONMENTAL MANAGEMENT PROJECT, IG-0496 SALE OF ENRICHED URANIUM AT...

66

CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix...

67

CRAD, Safety Basis - Y-12 Enriched Uranium Operations Oxide Conversion...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Safety Basis - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to...

68

CRAD, Management - Y-12 Enriched Uranium Operations Oxide Conversion...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE...

69

CRAD, Training - Y-12 Enriched Uranium Operations Oxide Conversion...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Training - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G...

70

CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of...

71

CRAD, Occupational Safety & Health - Y-12 Enriched Uranium Operations...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Occupational Safety & Health - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of...

72

CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to...

73

CRAD, Environmental Protection - Y-12 Enriched Uranium Operations...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Environmental Protection - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of...

74

CRAD, Radiological Controls - Y-12 Enriched Uranium Operations...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Radiological Controls - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of...

75

CRAD, Emergency Management - Y-12 Enriched Uranium Operations...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emergency Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Emergency Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility January...

76

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion  

E-Print Network (OSTI)

The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

77

Estimate of radiation release from MIT reactor with low enriched uranium (LEU) core during maximum hypothetical accident  

E-Print Network (OSTI)

In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis ...

Plumer, Kevin E. (Kevin Edward)

2011-01-01T23:59:59.000Z

78

Thermal breeder fuel enrichment zoning  

DOE Patents (OSTI)

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

Capossela, Harry J. (Schenectady, NY); Dwyer, Joseph R. (Albany, NY); Luce, Robert G. (Schenectady, NY); McCoy, Daniel F. (Latham, NY); Merriman, Floyd C. (Rotterdam, NY)

1992-01-01T23:59:59.000Z

79

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation  

SciTech Connect

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 use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

2007-10-01T23:59:59.000Z

80

Nickel container of highly-enriched uranium bodies and sodium  

SciTech Connect

A fuel element comprises highly a 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.

Zinn, Walter H. (Hinsdale, IL)

1976-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Disposition of excess highly enriched uranium status and update  

SciTech Connect

This paper presents the status of the US DOE program charged with the disposition of excess highly enriched uranium (HEU). Approximately 174 metric tonnes of HEU, with varying assays above 20 percent, has been declared excess from US nuclear weapons. A progress report on the identification and characterization of specific batches of excess HEU is provided, and plans for processing it into commercial nuclear fuel or low-level radioactive waste are described. The resultant quantities of low enriched fuel material expected from processing are given, as well as the estimated schedule for introducing the material into the commercial reactor fuel market. 2 figs., 3 tabs.

Williams, C.K. III; Arbital, J.G.

1997-09-01T23:59:59.000Z

82

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

Science Conference Proceedings (OSTI)

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.

Brim, Cornelia P.

2013-03-04T23:59:59.000Z

83

Overview of reduced enrichment fuels: Development, testing, and specification  

SciTech Connect

The US Reduced Enrichment Research and Test Reactor (RERTR) Program was established in 1978 to provide the technical means to operate research and test reactors with low enrichment uranium (LEU) fuels without significant penalty in experiment performance, operation costs, component modifications, or safety characteristics. This paper discusses relevant developments in fuel developments. 9 refs., 1 tab.

Snelgrove, J.L.

1987-01-01T23:59:59.000Z

84

SciTech Connect: enriched uranium  

Office of Scientific and Technical Information (OSTI)

enriched uranium Find enriched uranium Find How should I search Scitech Connect ... Basic or Advanced? Basic Search Advanced × Advanced Search Options Full Text: Bibliographic Data: Creator / Author: Name Name ORCID Title: Subject: Identifier Numbers: Research Org.: Sponsoring Org.: Site: All Alaska Power Administration, Juneau, Alaska (United States) Albany Research Center (ARC), Albany, OR (United States) Albuquerque Complex - NNSA Albuquerque Operations Office, Albuquerque, NM (United States) Amarillo National Resource Center for Plutonium, Amarillo, TX (United States) Ames Laboratory (AMES), Ames, IA (United States) Argonne National Laboratory (ANL), Argonne, IL (United States) Argonne National Laboratory-Advanced Photon Source (United States) Atlanta Regional Office, Atlanta, GA (United States) Atmospheric Radiation Measurement (ARM)

85

Recovery of Highly Enriched Uranium Provided to Foreign Countries...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Power Marketing Administration Other Agencies You are here Home Recovery of Highly Enriched Uranium Provided to Foreign Countries, DOEIG-0638 Recovery of Highly Enriched...

86

Table S1a. Uranium Purchased by Owners and Operators of U.S ...  

U.S. Energy Information Administration (EIA)

U.S. Energy Information Administration / 2010 Uranium Marketing ... deliveries, domestic, enrichers, enriched uranium, enrichment, fabricators, feed, foreign, fuel ...

87

The Office of Environmental Management Uranium Enrichment D&D...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Uranium Enrichment D&D The Office of Environmental Management Uranium Enrichment D&D Microsoft Word - B996F741.doc More Documents & Publications Microsoft Word - PSRP Updates...

88

DOE hands over uranium enrichment duties to government corporation  

SciTech Connect

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.

Simpson, J.

1993-07-15T23:59:59.000Z

89

Enrichment Determination of Uranium in Shielded Configurations  

Science Conference Proceedings (OSTI)

The determination of the enrichment of uranium is required in many safeguards and security applications. Typical methods of determining the enrichment rely on detecting the 186 keV gamma ray emitted by {sup 235}U. In some applications, the uranium is surrounded by external shields, and removal of the shields is undesirable. In these situations, methods relying on the detection of the 186 keV gamma fail because the gamma ray is shielded easily. Oak Ridge National Laboratory (ORNL) has previously measured the enrichment of shielded uranium metal using active neutron interrogation. The method consists of measuring the time distribution of fast neutrons from induced fissions with large plastic scintillator detectors. To determine the enrichment, the measurements are compared to a calibration surface that is created from Monte Carlo simulations where the enrichment in the models is varied. In previous measurements, the geometry was always known. ORNL is extending this method to situations where the geometry and materials present are not known in advance. In the new method, the interrogating neutrons are both time and directionally tagged, and an array of small plastic scintillators measures the uncollided interrogating neutrons. Therefore, the attenuation through the item along many different paths is known. By applying image reconstruction techniques, an image of the item is created which shows the position-dependent attenuation. The image permits estimating the geometry and materials present, and these estimates are used as input for the Monte Carlo simulations. As before, simulations predict the time distribution of induced fission neutrons for different enrichments. Matching the measured time distribution to the closest prediction from the simulations provides an estimate of the enrichment. This presentation discusses the method and provides results from recent simulations that show the importance of knowing the geometry and materials from the imaging system.

Crye, Jason Michael [ORNL; Hall, Howard L [ORNL; McConchie, Seth M [ORNL; Mihalczo, John T [ORNL; Pena, Kirsten E [ORNL

2011-01-01T23:59:59.000Z

90

US Department of Energy Uranium Enrichment Activity  

Science Conference Proceedings (OSTI)

KPMG Peat Marwick (KPMG), Certified Public Accountants, has completed its audit of the Department of Energy's Uranium Enrichment Activity (UEA) financial.statements as of September 30, 1991. The purpose of the audit was to determine whether (1) the financial statements were presented fairly in accordance with applicable accounting principles, (2) the auditee complied with all applicable laws and regulations that may have materially affected the financial statements, and (3) the internal accounting controls, taken as a whole, were adequate. The US Government, through the Department of Energy (DOE) and the management and operating contractor, operates the UEA to enrich uranium hexafluoride in the isotope U-235 for commercial power reactor operators, as further discussed in note 1 of the financial statements. The enrichment of uranium for Government program users, which had been a function of UEA, was transferred outside the UEA affective September 30, 1991, as described in note 3 of the financial statements. UEA is a part of DOE and does not exist as a separate legal entity. For financial reporting purposes, the entity is defined as those activities which provide enriching services to its customers. The financial statements are prepared by extracting and adjusting UEA related data from the financial records of DOE and its contractors.

Not Available

1992-06-16T23:59:59.000Z

91

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

SciTech Connect

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

Lillard, C. R.; Hayward, J. P.; Williamson, M. R.

2012-06-07T23:59:59.000Z

92

Possibility of nuclear pumped laser experiment using low enriched uranium  

SciTech Connect

Possibility to perform experiments for nuclear pumped laser oscillation by using low enriched uranium is investigated. Kinetic analyses are performed for two types of reactor design, one is using highly enriched uranium and the other is using low enriched uranium. The reactor design is based on the experiment reactor in IPPE. The results show the oscillation of nuclear pumped laser in the case of low enriched uranium reactor is also possible. The use of low enriched uranium in the experiment will make experiment easier.

Obara, Toru; Takezawa, Hiroki [Center for Research into Innovative Nuclear Energy Systems Tokyo Institute of Technology 2-12-1-N1-19, Ookayama Meguro-ku, Tokyo 152-8550 (Japan)

2012-06-06T23:59:59.000Z

93

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

SciTech Connect

Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

David W. Nigg; Sean R. Morrell

2012-09-01T23:59:59.000Z

94

Detection of uranium enrichment activities using environmental monitoring techniques  

SciTech Connect

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.

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

1993-03-30T23:59:59.000Z

95

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect

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

Travelli, A.

1982-01-01T23:59:59.000Z

96

High loading uranium fuel plate  

DOE Patents (OSTI)

Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pair of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat having a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process.

Wiencek, Thomas C. (Bolingbrook, IL); Domagala, Robert F. (Indian Head Park, IL); Thresh, Henry R. (Palos Heights, IL)

1990-01-01T23:59:59.000Z

97

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

SciTech Connect

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.

1995-07-05T23:59:59.000Z

98

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

E-Print Network (OSTI)

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

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

99

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel  

SciTech Connect

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.

A K Wertsching

2012-09-01T23:59:59.000Z

100

Dissolving uranium oxide--aluminum fuel  

SciTech Connect

The dissolution of aluminum-clad uranium oxide-aluminum fuel was studied to provide basic data for dissolving this type of enriched uranium fuel at the Savannah River Plant. The studies also included the dissolution of a similar material prepared from scrap uranium oxides that were to be recycled through the solvent extraction process. The dissolving behavior of uranium oxide-aluminum core material is similar to that of U-Al alloy. Dissolving rates are rapid in HNO/sub 3/-Hg(NO/sub 3/)/sub 2/ solutions. Irradiation reduce s the dissolving rate and increases mechanical strength. A dissolution model for use in nuclear safety analyses is developed, . based on the observed dissolving characteristics. (auth)

Perkins, W.C.

1973-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

A Lifecycle Emissions Model (LEM): Lifecycle Emissions from Transportation Fuels, Motor Vehicles, Transportation Modes, Electricity Use, Heating and Cooking Fuels, and Materials  

E-Print Network (OSTI)

provide the world’s uranium enrichment services. With theseenergy efficiency of uranium enrichment, in mWh-enrichment-gas Motor-vehicle flows Uranium enrichment Agriculture Fuel

Delucchi, Mark

2003-01-01T23:59:59.000Z

102

CRAD, Emergency Management - Y-12 Enriched Uranium Operations Oxide  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emergency Management - Y-12 Enriched Uranium Operations Oxide Emergency Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Emergency Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Emergency Management program at the Y-12 Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Emergency Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations Oxide

103

CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations Oxide  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Conduct of Operations - Y-12 Enriched Uranium Operations Conduct of Operations - Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January, 2005 assessment of Conduct of Operations program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion

104

CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Criticality Safety program at the Y-12 - Enriched Uranium Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Criticality Safety - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion

105

CRAD, Management - Y-12 Enriched Uranium Operations Oxide Conversion  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Management program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion

106

CRAD, Training - Y-12 Enriched Uranium Operations Oxide Conversion Facility  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Operations Oxide Conversion Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Training - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Training Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications CRAD, Conduct of Operations - Y-12 Enriched Uranium Operations Oxide

107

Y-12 and the super enriched Uranium 235?  

NLE Websites -- All DOE Office Websites (Extended Search)

"super enriched Uranium 235" Ken Bernander called me to say that he had read in the newspaper about the 100 milligrams of uranium oxide that is 99.999% U-235. He was chuckling when...

108

Candidate processes for diluting the {sup 235}U isotope in weapons-capable highly enriched uranium  

SciTech Connect

The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile {sup 235}U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile {sup 235}U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel.

Snider, J.D.

1996-02-01T23:59:59.000Z

109

Active interrogation of highly enriched uranium  

SciTech Connect

Active interrogation techniques provide reliable detection of highly enriched uranium (HEU) even when passive detection is difficult. We use 50-Hz pulsed beams of bremsstrahlung photons from a 10-MeV linac or 14-MeV neutrons from a neutron generator for interrogation, thus activating the HEU. Detection of neutrons between pulses is a positive indicator of the presence of fissionable material. We detect the neutrons with three neutron detector designs based on {sup 3}He tubes. This report shows examples of the responses in these three detectors, for unshielded and shielded kilogram quantities of HEU, in containers as large as cargo containers.

Moss, C. E. (Calvin E.); Hollas, C. L. (Charles L.); Myers, W. L. (William L.)

2004-01-01T23:59:59.000Z

110

Chapter 20 - Uranium Enrichment Decontamination & Decommissioning Fund  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

0. Uranium Enrichment Decontamination and Decommissioning Fund 20-1 0. Uranium Enrichment Decontamination and Decommissioning Fund 20-1 CHAPTER 20 URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND 1. INTRODUCTION. a. Purpose. To establish policies and procedures for the financial management, accounting, budget preparation, cash management of the Uranium Enrichment Decontamination and Decommissioning Fund, referred to hereafter as the Fund. b. Applicability. This chapter applies to all Departmental elements, including the National Nuclear Security Administration, and activities that are directly or indirectly involved with the Fund. c. Requirements and Sources of the Fund. (1) The Energy Policy Act of 1992 (EPACT) requires DOE to establish and administer the Fund. EPACT authorizes that the

111

Highly Enriched Uranium Materials Facility, Major Design Changes...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Field Sites Power Marketing Administration Other Agencies You are here Home Highly Enriched Uranium Materials Facility, Major Design Changes Late...Lessons Learned Report, NNSA,...

112

Reestablishment of Enriched Uranium Operations at the Y-12 National...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Administration Other Agencies You are here Home Reestablishment of Enriched Uranium Operations at the Y-12 National Security Complex, DOEIG-0640 Reestablishment of...

113

Toxic Substances Control Act Uranium Enrichment Federal Facilities...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

McCall, Jr. http:www.em.doe.govffaaortsca.html 4252001 Toxic Substances Control Act Uranium Enrichment Federal Facilities Compliance Agree.. Page 12 of 26 Deputy Assistant...

114

Toxic Substances Control Act Uranium Enrichment Federal Facilities...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement (TSCA-UE- FFCA), February 20, 1992 State Kentucky Agreement Type Compliance Agreement Legal...

115

Toxic Substances Control Act Uranium Enrichment Federal Facilities...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement (TSCA-UE- FFCA), February 20, 1992 State Ohio Agreement Type Compliance Agreement Legal...

116

Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National...  

NLE Websites -- All DOE Office Websites (Extended Search)

Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

117

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

Science Conference Proceedings (OSTI)

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.

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

2010-02-01T23:59:59.000Z

118

EPA Update: NESHAP Uranium Activities  

E-Print Network (OSTI)

measurements have been performed on high-enriched uranium (HEU) oxide fuel pins and depleted uranium metal

119

CRAD, Environmental Protection - Y-12 Enriched Uranium Operations Oxide  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Environmental Protection - Y-12 Enriched Uranium Operations Environmental Protection - Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Environmental Protection - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Environmental Compliance program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Environmental Protection - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications

120

CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE Oversight - Y-12 Enriched Uranium Operations Oxide DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a DOE independent oversight assessment of the Y-12 Site Office's programs for oversight of its contractors at the Y-12 Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, DOE Oversight - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Accelerating the Reduction of Excess Russian Highly Enriched Uranium  

SciTech Connect

This paper presents the latest information on one of the Accelerated Highly Enriched Uranium (HEU) Disposition initiatives that resulted from the May 2002 Summit meeting between Presidents George W. Bush and Vladimir V. Putin. These initiatives are meant to strengthen nuclear nonproliferation objectives by accelerating the disposition of nuclear weapons-useable materials. The HEU Transparency Implementation Program (TIP), within the National Nuclear Security Administration (NNSA) is working to implement one of the selected initiatives that would purchase excess Russian HEU (93% 235U) for use as fuel in U.S. research reactors over the next ten years. This will parallel efforts to convert the reactors' fuel core from HEU to low enriched uranium (LEU) material, where feasible. The paper will examine important aspects associated with the U.S. research reactor HEU purchase. In particular: (1) the establishment of specifications for the Russian HEU, and (2) transportation safeguard considerations for moving the HEU from the Mayak Production Facility in Ozersk, Russia, to the Y-12 National Security Complex in Oak Ridge, TN.

Benton, J; Wall, D; Parker, E; Rutkowski, E

2004-02-18T23:59:59.000Z

122

Disposition of Surplus Highly Enriched Uranium  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

@ @ Printed with soy ink on recycled paper. ,, ,, This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors horn the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; telephone (423) 576-8401 for prices, Available to the public from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. Copies of this document are available (while supplies last) upon written request to: Office of Fissile Materials Disposition, MD-4 ' Forrestal Building United States Department of Energy 1000 Independence Avenue, SW Washington, DC 20585 Department of Energy Washington, DC 20585 June 1996 Dear hterested Party: The Disposition of Surplus Highly Enriched Uranium Final Environmental Impact Statemnt is enclosed for your information. This document has been prepared in accordance

123

Uranium enrichment decontamination and decommissioning fund  

SciTech Connect

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.

1994-12-31T23:59:59.000Z

124

High-uranium-loaded U/sub 3/O/sub 8/--Al fuel element development program  

SciTech Connect

The High-Uranium-Loaded U/sub 3/O/sub 8/--Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages.

Martin, M.M.

1978-01-01T23:59:59.000Z

125

Conversion and Blending Facility highly enriched uranium to low enriched uranium as uranyl nitrate hexahydrate. Revision 1  

SciTech Connect

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. 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. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.

1995-07-05T23:59:59.000Z

126

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

Science Conference Proceedings (OSTI)

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.

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

2011-02-01T23:59:59.000Z

127

Tank 41H bounding uranium enrichment  

Science Conference Proceedings (OSTI)

The intent of this document is to combine data from salt samples and historical process information to bound the uranium (U-235) enrichment which could be expected in the upper portion of the salt in Tank 41H. This bounding enrichment will be used in another document to establish a nuclear safety basis for initial salt removal operations. During the processing period of interest (4/82-4/87), waste was fed to the 2H Evaporator from Tank 43H, and the evaporator bottoms were sent to Tank 41H where the bottoms were allowed to cool (resulting in the formation of salt deposits in the tank). As Tank 41H was filled with concentrate, the supernate left after salt formation was recycled back to Tank 43H and reprocessed through the evaporator along with any additional waste which had been added to Tank 43H. As Tank 41 H filled with salt, this recycle took place with increasing frequency because it took less time to fill the decreased volume with evaporator concentrate. By determining which of the sampled waste tanks were receiving fresh waste from the canyons at the time the tanks were sampled (from published transfer records), it was possible to deduce which samples were likely representative of fresh canyon waste. The processing that was being carried out in the Separation canyons when these tanks were sampled, should be comparable to the processing while Tank 41H was being filled.

Cavin, W.S.

1994-07-12T23:59:59.000Z

128

Development of a low enrichment uranium core for the MIT reactor.  

E-Print Network (OSTI)

??An investigation has been made into converting the MIT research reactor from using high enrichment uranium (HEU) to low enrichment uranium (LEU) with a newly… (more)

Newton, Thomas Henderson

2006-01-01T23:59:59.000Z

129

Low enrichment fuel conversion for Iowa State University  

SciTech Connect

This report discusses the UTR-10 reactor at Iowa State University which went critical on low enriched uranium (LEU) fuel on August 14, 1991. However, subsequent to the criticality experiments the fuel plates started to discolor. In addition, roll pins used to lift the fuel assemblies were discovered to be cracked. It was determined that these problems were due to chemical agents in the primary coolant water. The roll pins were replaced by solid stainless steel pins. The primary coolant was replaced and the reactor is currently in operation. Surveillance specimens will be used to monitor any possible future discoloration. The high enriched fuel (HEU) is being prepared for eventual shipment to a high enriched fuel receiving facility.

Rohach, A.F.

1992-08-01T23:59:59.000Z

130

Basic characterization of highly enriched uranium by gamma spectrometry  

E-Print Network (OSTI)

Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

Nguyen, C T

2006-01-01T23:59:59.000Z

131

Basic characterization of highly enriched uranium by gamma spectrometry  

E-Print Network (OSTI)

Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

Cong Tam Nguyen; Jozsef Zsigrai

2005-08-25T23:59:59.000Z

132

Selective Recovery of Enriched Uranium from Inorganic Wastes  

SciTech Connect

Uranium as U(IV) and U(VI) can be selectively recovered from liquids and sludge containing metal precipitates, inorganic salts, sand and silt fines, debris, other contaminants, and slimes, which are very difficult to de-water. Chemical processes such as fuel manufacturing and uranium mining generate enriched and natural uranium-bearing wastes. This patented Framatome ANP (FANP) uranium recovery process reduces uranium losses, significantly offsets waste disposal costs, produces a solid waste that meets mixed-waste disposal requirements, and does not generate metal-contaminated liquids. At the head end of the process is a floating dredge that retrieves liquids, sludge, and slimes in the form of a slurry directly from the floor of a lined surface impoundment (lagoon). The slurry is transferred to and mixed in a feed tank with a turbine mixer and re-circulated to further break down the particles and enhance dissolution of uranium. This process uses direct steam injection and sodium hypochlorite addition to oxidize and dissolves any U(IV). Cellulose is added as a non-reactive filter aid to help filter slimes by giving body to the slurry. The slurry is pumped into a large recessed-chamber filter press then de-watered by a pressure cycle-controlled double-diaphragm pump. U(VI) captured in the filtrate from this process is then precipitated by conversion to U(IV) in another Framatome ANP-patented process which uses a strong reducing agent to crystallize and settle the U(IV) product. The product is then dewatered in a small filter press. To-date, over 3,000 Kgs of U at 3% U-235 enrichment were recovered from a 8100 m2 hypalon-lined surface impoundment which contained about 10,220 m3 of liquids and about 757 m3 of sludge. A total of 2,175 drums (0.208 m3 or 55 gallon each) of solid mixed-wastes have been packaged, shipped, and disposed. In addition, 9463 m3 of low-U liquids at <0.001 KgU/m3 were also further processed and disposed.

Kimura, R. T.

2003-02-26T23:59:59.000Z

133

Transportable calorimeter measurements of highly enriched uranium  

SciTech Connect

A sensitive calorimeter has been combined with a small temperature-controlled water bath to compose a transportable system that is capable of measuring multikilogram quantities of highly enriched uranium (HEU). The sample chamber size, 5 in. in diameter by 10 in. high, is large enough to hold sufficient HEU metal or high-grade scrap to provide a measurable thermal signal. Calorimetric measurements performed on well-characterized material indicate that the thermal power generated by 93% {sup 235}U samples with 1.0% {sup 234}U can be measured with a precision of about 1% (1 sigma) for 4-kg samples. The transportable system consists of a twin-bridge calorimeter installed inside a 55-gal. stainless steel drum filled with water with heating and cooling supplied by a removable thermoelectric module attached to the side. Isotopic measurements using high-resolution gamma-ray measurements of the HEU samples and analysis with the FRAM code were used to determine the isotopic ratios and specific power of the samples. This information was used to transform the measured thermal power into grams of HEU. Because no physical standards are required, this system could be used for the verification of plutonium, {sup 238}Pu heat sources, or large quantities of metal or other high-grade matrix forms of HEU.

Rudy, C.; Bracken, D.S.; Staples, P.; Carrillo, L.

1997-11-01T23:59:59.000Z

134

Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear  

National Nuclear Security Administration (NNSA)

Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Oak Ridge, Tenn. Selected as Uranium Enrichment Site Oak Ridge, Tenn. Selected as Uranium Enrichment Site September 19, 1942 Oak Ridge, TN

135

Nuclear Fuel Facts: Uranium | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Uranium Management and Uranium Management and Policy » Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing minerals such as uraninite. Uranium ore can be mined from open pits or underground excavations. The ore can then be crushed and treated at a mill to separate the valuable uranium from the ore. Uranium may also be dissolved directly from the ore deposits

136

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

SciTech Connect

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)

1976-04-01T23:59:59.000Z

137

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network (OSTI)

measurement of uranium enrichment and actinide compositionsIE indicates initial enrichment of uranium in the fuel, M U

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

138

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

SciTech Connect

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.

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

2008-06-05T23:59:59.000Z

139

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

Science Conference Proceedings (OSTI)

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.

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

2009-03-01T23:59:59.000Z

140

Safeguarding a NWS International Enrichment Center as an Enriched Uranium Store  

SciTech Connect

The operational and regulatory singularities of a multilateral facility designed to provide enriched uranium to a consortium of members may engender a new sub-category of safeguard criteria for the International Atomic Energy Agency (IAEA). This paper introduces the contingency of monitoring such a facility as a uranium storage center with cylinders containing low-enriched uranium (LEU) as the principal, and perhaps only, material open to verification. Accountancy and verification techniques will be proffered together with disparate means for maintaining continuity of knowledge (CoK) on verified stock.

Curtis, Michael M.

2008-03-31T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

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

DOE Green Energy (OSTI)

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

Lichtenwalter, J.J.

1997-07-01T23:59:59.000Z

142

RERTR program reduces use of enriched uranium in research reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

RERTR program reduces use of enriched uranium in research reactors RERTR program reduces use of enriched uranium in research reactors worldwide Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share RERTR program reduces use of enriched uranium in research reactors worldwide The High Flux Reactor in Petten, the Netherlands READY TO CONVERT - The High Flux Reactor in Petten, the Netherlands, has

143

GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium |  

National Nuclear Security Administration (NNSA)

Program: Minimizing the Use of Highly Enriched Uranium | Program: Minimizing the Use of Highly Enriched Uranium | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > GTRI's Convert Program: Minimizing the Use of ... Fact Sheet GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium Apr 12, 2013

144

Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Materials & Waste » Nuclear Materials & Waste » Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a welded 3013 containers that are nested in 9975 shipping containers. 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a welded 3013 containers that are nested in 9975 shipping

145

SciTech Connect: "enriched uranium"  

Office of Scientific and Technical Information (OSTI)

enriched uranium" Find enriched uranium" Find How should I search Scitech Connect ... Basic or Advanced? Basic Search Advanced × Advanced Search Options Full Text: Bibliographic Data: Creator / Author: Name Name ORCID Title: Subject: Identifier Numbers: Research Org.: Sponsoring Org.: Site: All Alaska Power Administration, Juneau, Alaska (United States) Albany Research Center (ARC), Albany, OR (United States) Albuquerque Complex - NNSA Albuquerque Operations Office, Albuquerque, NM (United States) Amarillo National Resource Center for Plutonium, Amarillo, TX (United States) Ames Laboratory (AMES), Ames, IA (United States) Argonne National Laboratory (ANL), Argonne, IL (United States) Argonne National Laboratory-Advanced Photon Source (United States) Atlanta Regional Office, Atlanta, GA (United States) Atmospheric Radiation Measurement (ARM)

146

EIS-0240: Disposition of Surplus Highly Enriched Uranium  

Energy.gov (U.S. Department of Energy (DOE))

The Department proposes to eliminate the proliferation threat of surplus highly enriched uranium (HEU) by blending it down to low enriched uranium (LEU), which is not weapons-usable. The EIS assesses the disposition of a nominal 200 metric tons of surplus HEU. The Preferred Alternative is, where practical, to blend the material for use as LEU and use overtime, in commercial nuclear reactor field to recover its economic value. Material that cannot be economically recovered would be blended to LEU for disposal as low-level radioactive waste.

147

Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania  

SciTech Connect

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 for conversion to low enriched uranium. 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 Chelyabinsk, 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 Rosatom and the International Atomic Energy Agency. 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.

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

2010-03-01T23:59:59.000Z

148

Nuclear Isotopic Dilution of Highly-Enriched Uranium-235 and Uranium-233 by Dry Blending via the RM-2 Mill Technology  

SciTech Connect

The United States Department of Energy has initiated numerous activities to identify strategies to disposition various excess fissile materials. Two such materials are the off-specification highly enriched uranium-235 oxide powder and the uranium-233 contained in unirradiated nuclear fuel both currently stored at the Idaho National Engineering and Environmental Laboratory. This report describes the development of a technology that could dilute these materials to levels categorized as low-enriched uranium, or further dilute the materials to a level categorized as waste. This dilution technology opens additional pathways for the disposition of these excess fissile materials as existing processing infrastructure continues to be retired.

N. A. Chipman; R. N. Henry; R. K. Rajamani; S. Latchireddi; V. Devrani; H. Sethi; J. L. Malhotra

2004-02-01T23:59:59.000Z

149

ENRICHMENT DETERMINATION OF URANIUM METAL IN SHIELDED CONFIGURATIONS WITHOUT CALIBRATION STANDARDS.  

E-Print Network (OSTI)

??The determination of the enrichment of uranium is required in many safeguards and security applications. Typical methods to determine the enrichment rely on detecting the… (more)

Crye, Jason Michael

2013-01-01T23:59:59.000Z

150

IRRADIATION EFFECTS ON ZIRCONIUM-CLAD URANIUM-ZIRCONIUM FUEL PLATES  

SciTech Connect

This report summarizes the series of irradiations conducted in a Hanford reactor on specimens of zirconiumclad, uranium-- zirconium fuel plates containing 3, 6, and 14 vt.% highly (93.4%) enriched uranium. More than thirty fuel plates were exposed during the test program, which extended over a period of several years. (auth)

Bailey, R.E.

1958-02-01T23:59:59.000Z

151

Tank 41H bounding uranium enrichment. Revision 1  

Science Conference Proceedings (OSTI)

The intent of this document is to combine data from salt samples and historical process information to bound the uranium (U-235) enrichment which could be expected in the upper portion of the salt in Tank 41H. This bounding enrichment will be used in another document to establish a nuclear safety basis for initial salt removal operations. Any number of mixing scenarios could have been examined for the components which fed the evaporator during the formation of the last five feet of salt. The scenario presented was designed to be conservative, while still incorporating process knowledge and available data where possible. In the scenario, the lowest enrichment seen in any feed material was for the L4 feed which was evaporated to form the top part of the salt in Tank 41H. The lowest enrichment of 17% is still higher than the 16% (95% confidence) maximum enrichment actually found at the salt surface (from sample results). This leads to the conclusion that the uranium enrichment of the material (L1) which was being fed to the evaporate when the last five feet began to form, was lower than 22%. The conservatism used in this analysis, combined with the available sample data are believed to provide a defensible basis for establishing an upper bounding enrichment of 22% for the top five feet of salt.

Cavin, W.S.

1994-09-30T23:59:59.000Z

152

A confirmatory measurement technique for highly enriched uranium  

SciTech Connect

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 /sup 235/U, the 1001 keV originates from a daughter of /sup 238/U, and the 2614 keV originates from a daughter of /sup 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.

Sprinkle, J.K. Jr.

1987-07-01T23:59:59.000Z

153

Uranium Enrichment Standards of the Y-12 Nuclear Detection and Sensor Testing Center  

SciTech Connect

The Y-12 National Security Complex has recently fabricated and characterized a new series of metallic uranium standards for use in the Nuclear Detection and Sensor Testing Center (NDSTC). Ten uranium metal disks with enrichments varying from 0.2 to 93.2% {sup 235}U were designed to provide researchers access to a wide variety of measurement scenarios in a single testing venue. Special care was taken in the selection of the enrichments in order to closely bracket the definitions of reactor fuel at 4% {sup 235}U and that of highly enriched uranium (HEU) at 20% {sup 235}U. Each standard is well characterized using analytical chemistry as well as a series of gamma-ray spectrometry measurements. Gamma-ray spectra of these standards are being archived in a reference library for use by customers of the NDSTC. A software database tool has been created that allows for easier access and comparison of various spectra. Information provided through the database includes: raw count data (including background spectra), regions of interest (ROIs), and full width half maximum calculations. Input is being sought from the user community on future needs including enhancements to the spectral database and additional Uranium standards, shielding configurations and detector types. A related presentation are planned for the INMM 53rd Annual Meeting (Hull, et al.), which describe new uranium chemical compound standards and testing opportunities at Y-12 Nuclear Detection and Sensor Testing Center (NDSTC).

Cantrell, J.

2012-05-23T23:59:59.000Z

154

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion.  

E-Print Network (OSTI)

??The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density… (more)

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

155

Mechanical Characteristics of Fresh and Irradiated U-10Mo Fuel  

Science Conference Proceedings (OSTI)

Abstract Scope, The Reduced Enrichment for Research and Test Reactors ... test reactors to use low-enriched uranium instead of high-enriched uranium fuels ...

156

Measurement of uranium enrichment by gamma spectroscopy: result of an experimental design  

E-Print Network (OSTI)

PAPER Measurement of uranium enrichment by gamma spectroscopy: result of an experimental design Gamma spectroscopy is commonly used in nuclear safeguards to measure uranium enrichment. An experimental design has been carried out for the measurement of uranium enrichment using this technique with different

157

Moderation control in low enriched {sup 235}U uranium hexafluoride packaging operations and transportation  

Science Conference Proceedings (OSTI)

Moderation control is the basic parameter for ensuring nuclear criticality safety during the packaging and transport of low {sup 235}U enriched uranium hexafluoride before its conversion to nuclear power reactor fuel. Moderation control has permitted the shipment of bulk quantities in large cylinders instead of in many smaller cylinders and, therefore, has resulted in economies without compromising safety. Overall safety and uranium accountability have been enhanced through the use of the moderation control. This paper discusses moderation control and the operating procedures to ensure that moderation control is maintained during packaging operations and transportation.

Dyer, R.H. [USDOE Oak Ridge Operations Office, TN (United States); Kovac, F.M. [Oak Ridge National Lab., TN (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

1993-10-01T23:59:59.000Z

158

Critical masses of highly enriched uranium diluted with Gd and polyethylene  

SciTech Connect

A series of experiments have been performed containing highly enriched uranium, hydrogenous moderator (polyethylene), and gadolinium as a neutron absorber. The purpose of the experiments is to provide additional criticality data that can be used to verify and validate criticality safety evaluations in support of the National Spent Fuel Program. In addition, the experiments were also designed to provide criticality data for heterogeneous systems as noted in reference 1.

Sanchez, R. G. (Rene G.); Loaiza, D. J. (David J.); Bennion, J. (John)

2001-01-01T23:59:59.000Z

159

Unallocated Off-Specification Highly Enriched Uranium: Recommendations for Disposition  

Science Conference Proceedings (OSTI)

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.

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

2002-02-27T23:59:59.000Z

160

Report on the Effect the Low Enriched Uranium Delivered Under the Highly  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Report on the Effect the Low Enriched Uranium Delivered Under the Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion The successful implementation of the HEU Agreement remains a high priority of the U.S. Government. The agreement also serves U.S. and Russian commercial interests. HEU Agreement deliveries are an important source of supply in meeting requirements for U.S. utility uranium, conversion, and

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Report on the Effect the Low Enriched Uranium Delivered Under the Highly  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

on the Effect the Low Enriched Uranium Delivered Under the on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion The successful implementation of the HEU Agreement remains a high priority of the U.S. Government. The agreement also serves U.S. and Russian commercial interests. HEU Agreement deliveries are an important source of supply in meeting requirements for U.S. utility uranium, conversion, and

162

SOLUBLE POISONS FOR SLIGHTLY ENRICHED URANIUM SYSTEMS  

DOE Patents (OSTI)

A study of B and Th poisoning of slightly enriched U/sup 235/ hetcrogeneous and homogencous systems has been made. This study indicates large processing plant capacity increases are possible by the incorporation of soluble neutron poisons. A tabulation of other readily available neutron poisons together with their poisoning effects has been made. The importance of being able to remove the ncutron poisons when desired as well as having them present under all conditions where nuclear safety is dependent upon them has also been presented. (auth)

Ketzlach, N.

1957-05-01T23:59:59.000Z

163

Production and Handling Slide 1: The Uranium Fuel Cycle  

NLE Websites -- All DOE Office Websites (Extended Search)

and Handling The Uranium Fuel Cycle Skip Presentation Navigation Next Slide Last Presentation Table of Contents The Uranium Fuel Cycle Refer to caption below for image...

164

Simulation of transportation of low enriched uranium solutions  

SciTech Connect

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.

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

1996-08-01T23:59:59.000Z

165

Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor  

SciTech Connect

A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

Primm, Trent [ORNL; Gehin, Jess C [ORNL

2009-04-01T23:59:59.000Z

166

Continuing investigations for technology assessment of /sup 99/Mo production from LEU (low enriched Uranium) targets  

SciTech Connect

Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from /sup 99/Mo derived from the fissioning of high enriched uranium (HEU). The need for /sup 99m/Tc is continuing to grow, especially in developing countries, where needs and national priorities call for internal production of /sup 99/Mo. 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 /sup 99/Mo. Improvements in the electrodeposition of thin films of uranium metal are reported. These improvements 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 hastaloy. A cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets is reported. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminum alloy or uranium aluminide dispersed fuel used in other current target designs will allow the substitution of LEU for HEU in these targets with equivalent /sup 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 both concerns can be handled and that substitution of LEU for HEU can be achieved.

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

1987-01-01T23:59:59.000Z

167

108 An Investigation of the Mechanical Properties of U-Mo Fuels  

Science Conference Proceedings (OSTI)

Abstract Scope, The goal of the Reduced Enrichment for Research and Test ... test reactors to use low-enriched uranium instead of high-enriched uranium fuels,

168

Use of Savannah River Site facilities for blend down of highly enriched uranium  

SciTech Connect

Westinghouse Savannah River Company was asked to assess the use of existing Savannah River Site (SRS) facilities for the conversion of highly enriched uranium (HEU) to low enriched uranium (LEU). The purpose was to eliminate the weapons potential for such material. Blending HEU with existing supplies of depleted uranium (DU) would produce material with less than 5% U-235 content for use in commercial nuclear reactors. The request indicated that as much as 500 to 1,000 MT of HEU would be available for conversion over a 20-year period. Existing facilities at the SRS are capable of producing LEU in the form of uranium trioxide (UO{sub 3}) powder, uranyl nitrate [UO{sub 2}(NO{sub 3}){sub 2}] solution, or metal. Additional processing, and additional facilities, would be required to convert the LEU to uranium dioxide (UO{sub 2}) or uranium hexafluoride (UF{sub 3}), the normal inputs for commercial fuel fabrication. This study`s scope does not include the cost for new conversion facilities. However, the low estimated cost per kilogram of blending HEU to LEU in SRS facilities indicates that even with fees for any additional conversion to UO{sub 2} or UF{sub 6}, blend-down would still provide a product significantly below the spot market price for LEU from traditional enrichment services. The body of the report develops a number of possible facility/process combinations for SRS. The primary conclusion of this study is that SRS has facilities available that are capable of satisfying the goals of a national program to blend HEU to below 5% U-235. This preliminary assessment concludes that several facility/process options appear cost-effective. Finally, SRS is a secure DOE site with all requisite security and safeguard programs, personnel skills, nuclear criticality safety controls, accountability programs, and supporting infrastructure to handle large quantities of special nuclear materials (SNM).

Bickford, W.E.; McKibben, J.M.

1994-02-01T23:59:59.000Z

169

Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor  

SciTech Connect

Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

1983-01-01T23:59:59.000Z

170

Enrichment and Location of Uranium Precipitates from Uranyl Carbonate Addition to Tank 43  

SciTech Connect

In order to safety restart the 2H evaporator, plans are to add depleted uranium (DU) as uranyl carbonate to Tank 43 to lower the 235U enrichment in the supernate. This memo examines the enrichment and location of uranium precipitates formed in Tank 43. An assessment of the risks associated with precipitating uranium shows that there is no criticality concern during this operation.

d' Entremont, P.D.

2001-06-04T23:59:59.000Z

171

Verification of Uranium Mass and Enrichments of Highly Enriched Uranium (HEU) Using the Nuclear Materials Identification System (NMIS)  

SciTech Connect

This paper describes how the Nuclear Materials Identification System (NMIS), developed by the Oak Ridge National Laboratory (ORNL) and the Oak Ridge Y-12 Plant, was used to verify the mass and enrichment of hundreds of Highly Enriched Uranium (HEU) metal items in storage at the Y-12 Plant. The verifications had a relative spread of {+-}5% (3 sigma) with relative mean deviations from their declared values of +0.2% for mass and {minus}0.2% for enrichment. NMIS's capability to perform quantification of HEU enabled the Y-12 Plant to meet their nuclear material control and accountability (NMC and A) requirements. These verifications were performed in the storage vault in a very time and cost effective manner with as many as 55 verifications in one shift of operation.

Chiang, L.G.; Mattingly, J.K.; Ramsey, J.A.; Mihalczo, J.T.

2000-04-07T23:59:59.000Z

172

uranium hexafluoride - U.S. Energy Information Administration (EIA)  

U.S. Energy Information Administration (EIA)

Uranium fuel, nuclear reactors, generation, spent fuel. Total Energy. ... UF 6 is the form of uranium required for the enrichment process. Thank You.

173

AN EVALUATION OF THE URANIUM CONTAMINATION ON THE SURFACES OF ALCLAD URANIUM-ALUMINUM ALLOY RESEARCH REACTOR FUEL PLATES  

SciTech Connect

Reported radioactivity in the Low-Intensity Test Reactor (LITR) water coolant traceable to uranium contamination on the surfaces of the alclad uranium-- aluminum plate-tyne fuel element led to an investigation to determine the sources of uranium contamination on the fuel plate surfaces. Two possible contributors to surface contamination are external sources such as rolling-mill equipment, the most obvious, and diffusion of uranium from the uranium-aluminum alloy fuel into the aluminum cladding. This diffusion is likely because of the 600 deg C heat treatments used in the conventional fabrication process. Uranium determinations based on neutron activation analysis of machined layers from fuel plate surfaces showed that rolling-mill equipment, contaminated with highly enriched uranium, was responsible for transferring as much as 180 ppm U to plate surfaces. By careful practice where cleanliness is emphasized, surface contamination can be reduced to 0.6 ppm U/sup 235/. The residue remaining on the plate surface may be accounted for by diffusion of uranium from the fuel alloy into and through the cladding of the fuel plate. Data obtained from preliminary diffusion studies permitted a good estimate to be made of the diffusion coefficient of uranium into aluminum at 600 deg C: 2.5 x 10/sup -8/ cm//sec. To minimize diffusion while the plate-type aluminum-base research reactor fuel element is being processed, heat treatments at 600 deg C should be limited to 2.5 hr. The uranium contamination on the surfaces of the finished fuel plates should then be less than 0.6 ppm U / sup 235/ . This investigation also revealed that the solubility limit of uranium in aluminum at 600 deg C is approx 60 ppm. (auth)

Beaver, R.J.; Erwin, J.H.; Mateer, R.S.

1962-03-19T23:59:59.000Z

174

Fabrication of Uranium Oxycarbide Kernels for HTR Fuel  

Science Conference Proceedings (OSTI)

Babcock and Wilcox (B&W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-µm, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B&W produced 425-µm, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B&W also produced 500-µm, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B&W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

Charles Barnes; CLay Richardson; Scott Nagley; John Hunn; Eric Shaber

2010-10-01T23:59:59.000Z

175

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

SciTech Connect

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 {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

Travelli, A.

1994-05-01T23:59:59.000Z

176

Photon and neutron active interrogation of highly enriched uranium.  

SciTech Connect

The physics of photon and neutron active interrogation of highly enriched uranium (HEU) using the delayed neutron reinterrogation method is described in this paper. Two sets of active interrogation experiments were performed using a set of subcritical configurations of cocentric HEU metal hemishells. One set of measurements utilized a pulsed 14-MeV neutron generator as the active source. The second set of measurements utilized a linear accelerator-based bremsstrahlung photon source as an active interrogation source. The neutron responses were measured for both sets of experiments. The operational details and results for both measurement sets are described.

Myers, W. L. (William L.); Goulding, C. A. (Charles A.); Hollas, C. L. (Charles L.); Moss, C. E. (Calvin E.)

2004-01-01T23:59:59.000Z

177

ESTIMATING HISTORICAL TRICHLOROETHYLENE EXPOSURE IN A URANIUM ENRICHMENT, GASEOUS DIFFUSION PLANT.  

E-Print Network (OSTI)

??Previous studies at two uranium enrichment plants have looked at radiation exposures, but not an extensive list of chemical exposures, limiting evaluation of potential interactions.… (more)

MOSER, ADRIANE

2005-01-01T23:59:59.000Z

179

Production and Handling Slide 5: The Uranium Fuel Cycle  

NLE Websites -- All DOE Office Websites (Extended Search)

Refer to caption below for image description The third step in the uranium fuel cycle involves the conversion of "yellowcake" to uranium hexafluoride (UF6), the chemical form...

180

Over 90% of uranium purchased by U.S. commercial nuclear reactors ...  

U.S. Energy Information Administration (EIA)

Uranium fuel, nuclear reactors ... and enrichment. EIA's 2010 Uranium Marketing Annual Report presents data on purchases and sales of uranium contracts and ...

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

FUEL CYCLE COSTS IN A GRAPHITE MODERATED SLIGHTLY ENRICHED FUSED SALT REACTOR  

SciTech Connect

A fuel cycle economic study has been made for a 315Mwe graphite- moderated slightly enriched fused-salt reactor. Fuel cycle costs of less than 1.5 mills may be possible for such reactors operating on a ten-year cycle even when the fuel is discarded at the end of the cycle. Recovery of the uranium and plutonium at the end of the cycle reduces the fuel cycle costs to approximates 1 mill/kwh. Changes in the waste storage cost, reprocessing cost or salt inventory have a relatively minor effect on fuel cycle costs. (auth)

Guthrie, C.E.

1959-01-01T23:59:59.000Z

182

Uranium mineralization in fluorine-enriched volcanic rocks  

Science Conference Proceedings (OSTI)

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).

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

1980-09-01T23:59:59.000Z

183

Initial report on characterization of excess highly enriched uranium  

SciTech Connect

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.

1996-07-01T23:59:59.000Z

184

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

SciTech Connect

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, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

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

2007-11-01T23:59:59.000Z

185

Fuel cycle optimization of thorium and uranium fueled PWR systems  

E-Print Network (OSTI)

The burnup neutronics of uniform PWR lattices are examined with respect to reduction of uranium ore requirements with an emphasis on variation of the fuel-to-moderator ratio

Garel, Keith Courtnay

1977-01-01T23:59:59.000Z

186

Progress in alkaline peroxide dissolution of low-enriched uranium metal and silicide targets  

SciTech Connect

This paper reports recent progress on two alkaline peroxide dissolution processes: the dissolution of low-enriched uranium metal and silicide (U{sub 3}Si{sub 2}) targets. These processes are being developed to substitute low-enriched for high-enriched uranium in targets used for production of fission-product {sup 99}Mo. Issues that are addressed include (1) dissolution kinetics of silicide targets, (2) {sup 99}Mo lost during aluminum dissolution, (3) modeling of hydrogen peroxide consumption, (4) optimization of the uranium foil dissolution process, and (5) selection of uranium foil barrier materials. Future work associated with these two processes is also briefly discussed.

Chen, L.; Dong, D.; Buchholz, B.A.; Vandegrift, G.F. [Argonne National Lab., IL (United States). Chemical Technology Div.; Wu, D. [Univ. of Illinois, Urbana, IL (United States)

1996-12-31T23:59:59.000Z

187

CRAD, Occupational Safety & Health - Y-12 Enriched Uranium Operations Oxide  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Y-12 Enriched Uranium Y-12 Enriched Uranium Operations Oxide Conversion Facility CRAD, Occupational Safety & Health - Y-12 Enriched Uranium Operations Oxide Conversion Facility January 2005 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Industrial Safety and Industrial Health programs at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Occupational Safety & Health - Y-12 Enriched Uranium Operations Oxide Conversion Facility More Documents & Publications

188

Highly enriched uranium (HEU) storage and disposition program plan  

SciTech Connect

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.

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

1995-01-01T23:59:59.000Z

189

Validation of NCSSHP for highly enriched uranium systems containing beryllium  

Science Conference Proceedings (OSTI)

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.

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

1994-09-29T23:59:59.000Z

190

Nuclear Fuel Materials - Programmaster.org  

Science Conference Proceedings (OSTI)

Mar 3, 2011 ... Various metallic nuclear fuels are bcc alloys of uranium that swell under ... that currently employ fuels containing highly enriched uranium.

191

Production and Handling Slide 3: The Uranium Fuel Cycle  

NLE Websites -- All DOE Office Websites (Extended Search)

First Slide Previous Slide Next Slide Last Presentation Table of Contents The Uranium Fuel Cycle See caption below for image description The second step in the uranium fuel cycle...

192

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

Science Conference Proceedings (OSTI)

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.

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

2011-05-12T23:59:59.000Z

193

Determination of the 235U enrichment of bulk uranium samples using delayed neutrons.  

SciTech Connect

A technique for utilizing the physics of the delayed neutron re-interrogation method to determine uranium enrichment is presented in this paper. A series of active interrogation measurements was performed using pulsed 14-MeV neutrons and a polyethylene moderated {sup 3}He based neutron detection system. Proof of principle measurements were performed on a set of bulk uranium oxide standards of differing enrichments. A series of measurements was performed on a set of uranium 'unknowns' with and without high-Z gamma-ray shielding (lead) present. Uranium enrichment estimates were obtained for all cases including the bulk uranium samples shielded by lead. Further refinement of this technique is needed to make it a more powerful tool for non-destructive assay of bulk uranium samples.

Myers, W. L. (William L.); Goulding, C. A. (Charles A.); Hollas, C. L. (Charles L.)

2006-01-01T23:59:59.000Z

194

Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core  

SciTech Connect

A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

Sterbentz, James W

2007-05-01T23:59:59.000Z

195

MOBILE SYSTEMS FOR DILUTION OF HIGHLY ENRICHED URANIUM AND URANIUM CONTAINING COMPONENTS  

SciTech Connect

A mobile melt-dilute (MMD) module for the treatment of aluminum research reactor spent fuel is being developed. The process utilizes a closed system approach to retain fission products/gases inside a sealed canister after treatment. The MMD process melts and dilutes spent fuel with depleted uranium to obtain a fissile fraction of less than 0.2. The final ingot is solidified inside the sealed canister and can be stored safely either wet or dry until final disposition or reprocessing. The MMD module can be staged at or near the research reactor fuel storage sites to facilitate the melt-dilute treatment of the spent fuel into a stable non-proliferable form.

Adams, T

2007-05-02T23:59:59.000Z

196

Fast critical assembly safeguards: NDA methods for highly enriched uranium. Summary report, October 1978-September 1979  

SciTech Connect

Nondestructive assay (NDA) methods, principally passive gamma measurements and active neutron interrogation, have been studied for their safeguards effectiveness and programmatic impact as tools for making inventories of highly enriched uranium fast critical assembly fuel plates. It was concluded that no NDA method is the sole answer to the safeguards problem, that each of those emphasized here has its place in an integrated safeguards system, and that each has minimum facility impact. It was found that the 185-keV area, as determined with a NaI detector, was independent of highly-enriched uranium (HEU) plate irradiation history, though the random neutron driver methods used here did not permit accurate assay of irradiated plates. Containment procedures most effective for accurate assaying were considered, and a particular geometry is recommended for active interrogation by a random driver. A model, pertinent to that geometry, which relates the effects of multiplication and self-absorption, is described. Probabilities of failing to detect that plates are missing are examined.

Bellinger, F.O.; Winslow, G.H.

1980-12-01T23:59:59.000Z

197

Realities of verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants  

SciTech Connect

Over a two and one-half year period beginning in 1981, representatives of six countries (United States, United Kingdom, Federal Republic of Germany, Australia, The Netherlands, and Japan) and the inspectorate organizations of the International Atomic Energy Agency and EURATOM developed and agreed to a technically sound approach for verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants. This effort, known as the Hexapartite Safeguards Project (HSP), led to the first international concensus on techniques and requirements for effective verification of the absence of weapons-grade nuclear materials production. Since that agreement, research and development has continued on the radiation detection technology-based technique that technically confirms the HSP goal is achievable. However, the realities of achieving the HSP goal of effective technical verification have not yet been fully attained. Issues such as design and operating conditions unique to each gas centrifuge plant, concern about the potential for sensitive technology disclosures, and on-site support requirements have hindered full implementation and operator support of the HSP agreement. In future arms control treaties that may limit or monitor fissile material production, the negotiators must recognize and account for the realities and practicalities in verifying the absence of HEU production. This paper will describe the experiences and realities of trying to achieve the goal of developing and implementing an effective approach for verifying the absence of HEU production. 3 figs.

Swindle, D.W.

1990-03-01T23:59:59.000Z

198

Critical masses of highly enriched uranium diluted with matrix material.  

SciTech Connect

Radioactive waste containing fissile material is frequently encountered in decontamination and decommissioning activities. For the most part, this waste is placed in containers or drums and stored in storage facilities. The amount of fissile material in each drum is generally small because of criticality safety limits that have been calculated with computer transport codes such as MCNP,1 KENO,2 or ONEDANT.3 To the best of our knowledge, no experimental critical mass data are available to verify the accuracy of these calculations or any calculations for systems containing fissile material (U-235, Pu-239, U-233) in contact with matrix material such as Al2O3, CaO, SiO2, Al, MgO, etc. The experiments presented in this paper establish the critical masses of highly enriched uranium foils diluted to various X/235U ratios with polyethylene and SiO2, polyethylene and aluminum, polyethylene and MgO, polyethylene and Gd, polyethylene and Fe, and moderated and reflected with polyethylene. In addition, these critical mass experimental data will be used to validate cross section data.

Sanchez, R. G. (Rene G.); Loaiza, D. J. (David J.); Kimpland, R. H. (Robert H.)

2002-01-01T23:59:59.000Z

199

MULTIPLICATION MEASUREMENTS WITH HIGHLY ENRICHED URANIUM METAL SLABS  

SciTech Connect

A series of neutron multiplication measurements with arrays of 1 by 8 by 10 in. slabs of 93.4% U/sup 235/-enriched uranium metal was made to provide data from which safety criteria for the storage of these fissile units can be established. Each slab contained 22.9 kg of U/sup 235/. A maximum of 125 units was assembled. The arrays studied were cubic lattices of the units and were usually parallelepipedal in shape. Arrays were both unmoderated and Plexiglas- moderated and were surrounded in most cases by a 1-in.-thick Plexiglas reflector. The lattice densities (ratio of fissile unit volume to lattice cell volume) were between 0.023 and 0.06. Unmoderated lattices with a density of 0.06 would require 145 plus or minus 5 units for criticality, while those with a density of 0.023 would require 350 plus or minus 30 units. In lattices in which the fissile units are separated by 1 in. of Plexiglas, approximately 27 units would be required for a critical array with a lattice density of 0.06 and about 75 units for a density of 0.023. Distributing Foamglas (containing 2% boron) throughout a moderated array increased the critical number of fissile units by a factor of 5, while Styrofoam had a small effect. (auth)

Mihalczo, J.T.; Lynn, J.J.

1959-07-27T23:59:59.000Z

200

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

Science Conference Proceedings (OSTI)

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.

Not Available

1987-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

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 %.  

SciTech Connect

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 Assembly 6F (ZPR-3/6F), the final phase of the Assembly 6 program, simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 47 at.%. Approximately 81.4% of the total fissions in this assembly occur above 100 keV, approximately 18.6% 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. 7 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3/6F began in late December 1956, and the experimental measurements were performed in January 1957. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates, perforated aluminum plates and stainless steel plates loaded into aluminum drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of three columns of 0.125 in.-wide (3.175 mm) HEU plates, three columns of 0.125 in.-wide depleted uranium plates, nine columns of 0.125 in.-wide perforated aluminum plates and one column of stainless steel plates. The maximum length of each column of core material in a drawer was 9 in. (228.6 mm). Because of the goal to produce an approximately spherical core, core fuel and diluent column lengths generally varied between adjacent drawers and frequently within an individual drawer. The axial reflector consisted of depleted uranium plates and blocks loaded in the available space in the front (core) drawers, with the remainder loaded into back drawers behind the front drawers. The radial reflector consisted of blocks of depleted uranium loaded directly into the matrix tubes. The assembly geometry approximated a reflected sphere as closely as the square matrix tubes, the drawers and the shapes of fuel and diluent plates allowed. According to the logbook and loading records for ZPR-3/6F

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

2010-09-30T23:59:59.000Z

202

BLENDING LOW ENRICHED URANIUM WITH DEPLETED URANIUM TO CREATE A SOURCE MATERIAL ORE THAT CAN BE PROCESSED FOR THE RECOVERY OF YELLOWCAKE AT A CONVENTIONAL URANIUM MILL  

SciTech Connect

Throughout the United States Department of Energy (DOE) complex, there are a number of streams of low enriched uranium (LEU) that contain various trace contaminants. These surplus nuclear materials require processing in order to meet commercial fuel cycle specifications. To date, they have not been designated as waste for disposal at the DOE's Nevada Test Site (NTS). Currently, with no commercial outlet available, the DOE is evaluating treatment and disposal as the ultimate disposition path for these materials. This paper will describe an innovative program that will provide a solution to DOE that will allow disposition of these materials at a cost that will be competitive with treatment and disposal at the NTS, while at the same time recycling the material to recover a valuable energy resource (yellowcake) for reintroduction into the commercial nuclear fuel cycle. International Uranium (USA) Corporation (IUSA) and Nuclear Fuel Services, Inc. (NFS) have entered into a commercial relationship to pursue the development of this program. The program involves the design of a process and construction of a plant at NFS' site in Erwin, Tennessee, for the blending of contaminated LEU with depleted uranium (DU) to produce a uranium source material ore (USM Ore{trademark}). The USM Ore{trademark} will then be further processed at IUC's White Mesa Mill, located near Blanding, Utah, to produce conventional yellowcake, which can be delivered to conversion facilities, in the same manner as yellowcake that is produced from natural ores or other alternate feed materials. The primary source of feed for the business will be the significant sources of trace contaminated materials within the DOE complex. NFS has developed a dry blending process (DRYSM Process) to blend the surplus LEU material with DU at its Part 70 licensed facility, to produce USM Ore{trademark} with a U235 content within the range of U235 concentrations for source material. By reducing the U235 content to source material levels in this manner, the material will be suitable for processing at a conventional uranium mill under its existing Part 40 license to remove contaminants and enable the product to re-enter the commercial fuel cycle. The tailings from processing the USM Ore{trademark} at the mill will be permanently disposed of in the mill's tailings impoundment as 11e.(2) byproduct material. Blending LEU with DU to make a uranium source material ore that can be returned to the nuclear fuel cycle for processing to produce yellowcake, has never been accomplished before. This program will allow DOE to disposition its surplus LEU and DU in a cost effective manner, and at the same time provide for the recovery of valuable energy resources that would be lost through processing and disposal of the materials. This paper will discuss the nature of the surplus LEU and DU materials, the manner in which the LEU will be blended with DU to form a uranium source material ore, and the legal means by which this blending can be accomplished at a facility licensed under 10 CFR Part 70 to produce ore that can be processed at a conventional uranium mill licensed under 10 CFR Part 40.

Schutt, Stephen M.; Hochstein, Ron F.; Frydenlund, David C.; Thompson, Anthony J.

2003-02-27T23:59:59.000Z

203

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

SciTech Connect

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.

1998-03-31T23:59:59.000Z

204

DISSOLUTION OF URANIUM FUELS BY MONOOR DIFLUOROPHOSPHORIC ACID  

DOE Patents (OSTI)

A method of dissolving and separating uranium from a uranium matrix fuel element by dissolving the uraniumcontaining matrix in monofluorophosphoric acid and/or difluorophosphoric acid at temperatures ranging from 150 to 275 un. Concent 85% C, thereafter neutralizing the solution to precipitate uranium solids, and converting the solids to uranium hexafluoride by treatment with a halogen trifluoride is presented. (AEC)

Johnson, R.; Horn, F.L.; Strickland, G.

1963-05-01T23:59:59.000Z

205

DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile November 7, 2005 - 12:38pm Addthis Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, DC - Secretary of Energy Samuel W. Bodman today announced that the Department of Energy's (DOE) National Nuclear Security Administration (NNSA) will remove up to 200 metric tons (MT) of Highly Enriched Uranium (HEU), in the coming decades, from further use as fissile material in U.S. nuclear weapons and prepare this material for other uses. Secretary Bodman made this announcement while addressing the 2005 Carnegie International Nonproliferation Conference in Washington, DC.

206

DOE to Remove 200 Metric Tons of Highly Enriched Uranium from...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Other Agencies You are here Home DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly...

207

NNSA Highly Enriched Uranium Removal Featured on The Rachel Maddow Show |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Highly Enriched Uranium Removal Featured on The Rachel Maddow Highly Enriched Uranium Removal Featured on The Rachel Maddow Show NNSA Highly Enriched Uranium Removal Featured on The Rachel Maddow Show March 22, 2012 - 11:37am Addthis NNSA Administrator Thomas D’Agostino appeared live last night to break the news with Rachel Maddow that all remaining weapons-usable material has been successfully removed from Mexico. | Photo courtesy of the NNSA. NNSA Administrator Thomas D'Agostino appeared live last night to break the news with Rachel Maddow that all remaining weapons-usable material has been successfully removed from Mexico. | Photo courtesy of the NNSA. Michael Hess Michael Hess Former Digital Communications Specialist, Office of Public Affairs What's the difference between HEU and LEU? Highly enriched uranium (HEU) has a greater than 20 percent

208

EA-1123: Transfer of Normal and Low-Enriched Uranium Billets...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

evaluates the environmental impacts of the proposal to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium to the United...

209

Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE  

Science Conference Proceedings (OSTI)

In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

Ade, Brian J [ORNL; Gauld, Ian C [ORNL

2011-10-01T23:59:59.000Z

210

Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems  

SciTech Connect

The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors.

1994-12-01T23:59:59.000Z

211

DOE/EA-1607: Final Environmental Assessment for Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium (June 2009)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

μCi/cc microcuries per cubic centimeter μCi/cc microcuries per cubic centimeter MAP mitigation action plan MEI maximally exposed individual mg/kg milligrams per kilogram mrem millirem mSv millisievert MT metric ton MTCA Model Toxics Control Act MTU metric tons of uranium N/A not applicable Final Environmental Assessment: Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium vi NAAQS National Ambient Air Quality Standards NEF National Enrichment Facility NEPA National Environmental Policy Act NRC U.S. Nuclear Regulatory Commission NU natural uranium NUF 6 natural uranium hexafluoride pCi/g picocuries per gram PEIS programmatic environmental impact statement PM 2.5 particulate matter with a diameter of 2.5 microns or less PM 10 particulate matter with a diameter of 10 microns or less

212

A SODIUM COOLED, GRAPHITE MODERATED, LOW ENRICHMENT URANIUM REACTOR FOR THE PRODUCTION OF USEFUL POWER  

SciTech Connect

A design study is presented for a sodium cooled, graphite moderated power reactor utilizing low enrichment uranium fuel. The design is characterized by dependence on existing technology and the use of standard, or nearly standard, components. The reactor has a nominal rating of 167 thermal megawatts, and a plant comprising three such reactors for a total output of 500 thermal megawatts is described. Sodium in a secondary, non-radioactive, circulation system carries the heat to a steam generator at 910 deg F and is returned at 420 deg F. Steam conditions at the turbine throttle are 600 psig and 825 deg F. Cost of the complete reactor power plant, consisting of the three reactors and one 150- megawatt turbogenerator, is estimated to be approximately ,165,000. (auth)

Weisner, E.F. ed.

1954-09-15T23:59:59.000Z

213

Implementation of the United States-Russian Highly Enriched Uranium Agreement: Current Status & Prospects  

SciTech Connect

The National Nuclear Security Administration's (NNSA) Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP) monitors and provides assurance that Russian weapons-grade HEU is processed into low enriched uranium (LEU) under the transparency provisions of the 1993 United States (U.S.)-Russian HEU Purchase Agreement. Meeting the Agreement's transparency provisions is not just a program requirement; it is a legal requirement. The HEU Purchase Agreement requires transparency measures to be established to provide assurance that the nonproliferation objectives of the Agreement are met. The Transparency concept has evolved into a viable program that consists of complimentary elements that provide necessary assurances. The key elements include: (1) monitoring by technical experts; (2) independent measurements of enrichment and flow; (3) nuclear material accountability documents from Russian plants; and (4) comparison of transparency data with declared processing data. In the interest of protecting sensitive information, the monitoring is neither full time nor invasive. Thus, an element of trust is required regarding declared operations that are not observed. U.S. transparency monitoring data and independent instrument measurements are compared with plant accountability records and other declared processing data to provide assurance that the nonproliferation objectives of the 1993 Agreement are being met. Similarly, Russian monitoring of U. S. storage and fuel fabrication operations provides assurance to the Russians that the derived LEU is being used in accordance with the Agreement. The successful implementation of the Transparency program enables the receipt of Russian origin LEU into the United States. Implementation of the 1993 Agreement is proceeding on schedule, with the permanent elimination of over 8,700 warhead equivalents of HEU. The successful implementation of the Transparency program has taken place over the last 10 years and has provided the necessary nonproliferation assurances to the U. S. while developing an increasing level of trust and cooperation between the U. S. and Russian government agencies.

R.rutkowski, E; Armantrout, G; Mastal, E; Glaser, J; Benton, J

2004-07-27T23:59:59.000Z

214

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

SciTech Connect

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

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

1994-11-01T23:59:59.000Z

215

Active-Interrogation Measurements of Induced-Fission Neutrons from Low-Enriched Uranium  

Science Conference Proceedings (OSTI)

Protection and control of nuclear fuels is paramount for nuclear security and safeguards; therefore, it is important to develop fast and robust controlling mechanisms to ensure the safety of nuclear fuels. Through both passive- and active-interrogation methods we can use fast-neutron detection to perform real-time measurements of fission neutrons for process monitoring. Active interrogation allows us to use different ranges of incident neutron energy to probe for different isotopes of uranium. With fast-neutron detectors, such as organic liquid scintillation detectors, we can detect the induced-fission neutrons and photons and work towards quantifying a sample’s mass and enrichment. Using MCNPX-PoliMi, a system was designed to measure induced-fission neutrons from U-235 and U-238. Measurements were then performed in the summer of 2010 at the Joint Research Centre in Ispra, Italy. Fissions were induced with an associated particle D-T generator and an isotopic Am-Li source. The fission neutrons, as well as neutrons from (n, 2n) and (n, 3n) reactions, were measured with five 5” by 5” EJ-309 organic liquid scintillators. The D-T neutron generator was available as part of a measurement campaign in place by Padova University. The measurement and data-acquisition systems were developed at the University of Michigan utilizing a CAEN V1720 digitizer and pulse-shape discrimination algorithms to differentiate neutron and photon detections. Low-enriched uranium samples of varying mass and enrichment were interrogated. Acquired time-of-flight curves and cross-correlation curves are currently analyzed to draw relationships between detected neutrons and sample mass and enrichment. In the full paper, the promise of active-interrogation measurements and fast-neutron detection will be assessed through the example of this proof-of-concept measurement campaign. Additionally, MCNPX-PoliMi simulation results will be compared to the measured data to validate the MCNPX-PoliMi code when used for active-interrogation simulations.

J. L. Dolan; M. J. Marcath; M. Flaska; S. A. Pozzi; D. L. Chichester; A. Tomanin; P. Peerani; G. Nebbia

2012-07-01T23:59:59.000Z

216

Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel  

Science Conference Proceedings (OSTI)

This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

DelCul, Guillermo D [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

2009-02-01T23:59:59.000Z

217

Estimated effect of eliminating TVA electricity demand charges on the price of enriched uranium  

Science Conference Proceedings (OSTI)

An estimate of the price of enrichment services from fiscal years 1984 through 1995 are forecast assuming demand charges were eliminated and TVA power rates were set. Uranium enrichment program officials estimated the TVA power rate and TVA officials confirmed the reasonableness of that estimate.

Not Available

1983-10-11T23:59:59.000Z

218

Dry Blending to Achieve Isotopic Dilution of Highly Enriched Uranium Oxide Materials  

SciTech Connect

The end of the cold war produced large amounts of excess fissile materials in the United States and Russia. The Department of Energy has initiated numerous activities to focus on identifying material management strategies for disposition of these excess materials. To date, many of these planning strategies have included isotopic dilution of highly enriched uranium as a means of reducing the proliferation and safety risks. Isotopic dilution by dry blending highly enriched uranium with natural and/or depleted uranium has been identified as one non-aqueous method to achieve these risk (proliferation and criticality safety) reductions. This paper reviews the technology of dry blending as applied to free flowing oxide materials.

Henry, Roger Neil; Chipman, Nathan Alan; Rajamani, R. K.

2001-04-01T23:59:59.000Z

219

MECHANISMS AND KINETICS OF URANIUM CORROSION AND URANIUM CORE FUEL ELEMENT RUPTURES IN WATER AND STEAM  

DOE Green Energy (OSTI)

The mechanisms and kinetics of uranium corrosion and fuel element ruptures were investigated in water and steam at 170 to 500 deg C and at 100 to 2800 psig. The fuel element samples were coextruded Zircaloy-clad uranium-core rods and tubes which were defected prior to exposure. Uranium corrosion was found to be the sum of two processes; direct oxidation by water, and oxidation of uranium hydride intermediate. Fuel element ruptures occur in two stages; an initial induction period followed by an accelerating corrosion of the core causing the cladding to blister, swell, and fracture. Uranium corrosion and fuel element ruptures were examined with respect to temperature, pressure, steam versus liquid water, heat treatment, carbon content of uranium, zirconium content of uranium, cladding thickness, fuel geometry, annular spacings, defect geometry and size, coolant flow, hydriding of Zircaloy components, and irradiation effects. (auth)

Troutner, V.H.

1960-07-21T23:59:59.000Z

220

Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States and the Government of the Russian Federation has on the  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Report on the Effect the Low Enriched Uranium Delivered Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States of America and the Government of the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Operation of the Gaseous Diffusion Plant 2008 Information Date: December 31, 2008 1 Introduction The Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning the Disposition of Highly Enriched Uranium Extracted from Nuclear Weapons (HEU Agreement) was signed on February 18, 1993. The HEU Agreement provides for the purchase over a 20-year period (1994-2013) of 500 metric tons (MT) of weapons-origin highly enriched uranium (HEU) from the Russian Federation

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221

Criteria for the safe storage of enriched uranium at the Y-12 Plant  

SciTech Connect

Uranium storage practices at US Department of Energy (DOE) facilities have evolved over a period spanning five decades of programmatic work in support of the nuclear deterrent mission. During this period, the Y-12 Plant in Oak Ridge, Tennessee has served as the principal enriched uranium facility for fabrication, chemical processing, metallurgical processing and storage. Recent curtailment of new nuclear weapons production and stockpile reduction has created significant amounts of enriched uranium available as a strategic resource which must be properly and safely stored. This standard specifies criteria associated with the safe storage of enriched uranium at the Y-12 Plant. Because programmatic needs, compliance regulations and desirable materials of construction change with time, it is recommended that these standards be reviewed and amended periodically to ensure that they continue to serve their intended purpose.

Cox, S.O.

1995-07-01T23:59:59.000Z

222

Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors  

Science Conference Proceedings (OSTI)

High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

Michael A. Pope

2012-07-01T23:59:59.000Z

223

Criticality of a Neptunium-237 sphere surrounded with highly enriched uranium shells and an iron reflector  

SciTech Connect

An additional experiment has been performed using the recently cast 6-kg {sup 237}Np sphere. The experiment consisted of surrounding the neptunium sphere with highly enriched uranium and an iron reflector. The purpose of the critical experiment is to provide additional criticality data that can be used to validate criticality safety evaluations involving the deposition of neptunium. It is well known that {sup 237}Np is primarily produced by successive neutron capture events in {sup 235}U or through the (n, 2n) reaction in {sup 238}U. These nuclear reactions lead to the production of {sup 237}U, which decays by beta emission into {sup 237}Np. In addition, in the spent fuel, {sup 241}Am decays by alpha emission into {sup 237}Np. Because {sup 237}Np is a threshold fissioner, the best reflectors for critical systems containing neptunium are those materials that exhibit good neutron scattering properties such as low carbon steel (99 wt % Fe). In this experiment, the iron reflector reduced the amount of uranium used in the critical experiment and increased the importance of the neptunium sphere.

Sanchez, R. G. (Rene G.); Loaiza, D. J. (David J.); Hayes, D. K. (David K.); Kimpland, R. H. (Robert H.)

2004-01-01T23:59:59.000Z

224

Effect of Highly Enriched/Highly Burnt UO2 Fuels on Fuel Cycle Costs, Radiotoxicity, and Nuclear Design Parameters  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Increased Enrichment/High Burnup and Light Water Reactor Fuel Cycle Optimization

Robert Gregg; Andrew Worrall

225

Environmental monitoring for detection of uranium enrichment operations: Comparison of LEU and HEU facilities  

SciTech Connect

In 1994, the International Atomic Energy Agency (IAEA) initiated an ambitious program of worldwide field trials to evaluate the utility of environmental monitoring for safeguards. Part of this program involved two extensive United States field trials conducted at the large uranium enrichment facilities. The Paducah operation involves a large low-enriched uranium (LEU) gaseous diffusion plant while the Portsmouth facilities include a large gaseous diffusion plant that has produced both LEU and high-enriched uranium (HEU) as well as an LEU centrifuge facility. As a result of the Energy Policy Act of 1992, management of the uranium enrichment operations was assumed by the US Enrichment Corporation (USEC). The facilities are operated under contract by Martin Marietta Utility Services. Martin Marietta Energy Systems manages the environmental restoration and waste management programs at Portsmouth and Paducah for DOE. These field trials were conducted. Samples included swipes from inside and outside process buildings, vegetation and soil samples taken from locations up to 8 km from main sites, and hydrologic samples taken on the sites and at varying distances from the sites. Analytical results from bulk analysis were obtained using high abundance sensitivity thermal ionization mm spectrometers (TIMS). Uranium isotopics altered from the normal background percentages were found for all the sample types listed above, even on vegetation 5 km from one of the enrichment facilities. The results from these field trials demonstrate that dilution by natural background uranium does not remove from environmental samples the distinctive signatures that are characteristic of enrichment operations. Data from swipe samples taken within the enrichment facilities were particularly revealing. Particulate analysis of these swipes provided a detailed ``history`` of both facilities, including the assays of the end product and tails for both facilities.

Hembree, D.M. Jr.; Carter, J.A.; Ross, H.H.

1995-03-01T23:59:59.000Z

226

Updated Uranium Fuel Cycle Environmental Impacts for Advanced Reactor Designs  

Science Conference Proceedings (OSTI)

The purpose of this project was to update the environmental impacts from the uranium fuel cycle for select advanced (GEN III+) reactor designs.

Nitschke, R.

2004-10-03T23:59:59.000Z

227

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

DOE Patents (OSTI)

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

Steinberg, Meyer (Huntington Station, NY); Powell, James R. (Shoreham, NY); Takahashi, Hiroshi (Setauket, NY); Grand, Pierre (Blue Point, NY); Kouts, Herbert (Brookhaven, NY)

1982-01-01T23:59:59.000Z

228

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel  

Science Conference Proceedings (OSTI)

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.

Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

2012-07-30T23:59:59.000Z

229

CONVERSION RATIOS IN SLIGHTLY ENRICHED URANIUM, WATER MODERATED LATTICES  

SciTech Connect

An experiment is described in which the conversion ratios were measured using highly enriched U-Al foils as catchers. Data are included on the ratios of epi-cadmium to sub-cadmium fission rates of U/sup 235/ in l% enriched U light water moderated lattices, and on conversion ratios of 1% enriched U light water moderated lattices. (J.R.D.)

Tassan, S.

1963-10-31T23:59:59.000Z

230

Advanced Neutron Source enrichment study  

SciTech Connect

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

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

1994-12-31T23:59:59.000Z

231

Risk Perceptions of Adults in the Town of Unicoi, Tennessee, Regarding the Possible Building of a Uranium Enrichment Plant.  

E-Print Network (OSTI)

??A prolonged siting controversy for a uranium enrichment facility has occurred in the Town of Unicoi, Tennessee. One hundred-seventy residents of Unicoi were interviewed using… (more)

Sellards, Shannon Kathleen

2004-01-01T23:59:59.000Z

232

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

Science Conference Proceedings (OSTI)

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.

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)

1992-07-01T23:59:59.000Z

233

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

DOE Patents (OSTI)

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.

McLean, II, William (Oakland, CA); Miller, Philip E. (Livermore, CA); Horton, James A. (Livermore, CA)

1995-01-01T23:59:59.000Z

234

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

DOE Patents (OSTI)

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.

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

1995-05-02T23:59:59.000Z

235

Environmental Assessment DOE/EA-1172 DOE Sale of Surplus Natural and Low Enriched Uranium  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE/EA-1172 DOE Sale of Surplus Natural and Low Enriched Uranium | October 1996 | For additional information contact: Office of Nuclear Energy, Science and Technology U.S. Department of Energy Washington, DC 20585 ii October 1996 | Table of Contents 1.0 Purpose and Need for Agency Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Purpose and Need for Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 Relationship to Other DOE NEPA Documents . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.2.1 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 . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 | 1.2.2 Disposition of Surplus Highly Enriched Uranium Final EIS . . . . . . . . 1-2 1.3 Public Comments on the Draft EA

236

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

Science Conference Proceedings (OSTI)

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

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

2012-07-01T23:59:59.000Z

237

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 %.  

Science Conference Proceedings (OSTI)

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 Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 12 began in late Jan. 1958, and the Assembly 12 program ended in Feb. 1958. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates and graphite plates loaded into stainless steel drawers which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, seven columns of 0.125 in.-wide depleted uranium plates and seven columns of 0.125 in.-wide graphite plates. The length of each column was 9 in. (228.6 mm) in each half of the core. The graphite plates were included to produce a softer neutron spectrum that would be more characteristic of a large power reactor. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the radial blanket was approximately 12 in. and the length of the radial blanket in each half of the matrix was 21 in. (533.4 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/12, the reference critical configuration was loading 10 which was critical on Feb. 5, 1958. The subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/12 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. An accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/12 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must d

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

2010-09-30T23:59:59.000Z

238

Uranium industry annual 1996  

SciTech Connect

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.

NONE

1997-04-01T23:59:59.000Z

239

Analysis of waste matrix material experiments mixed with highly enriched uranium on the thermal energy region  

SciTech Connect

The basic characteristics of waste materials such as silicon dioxide, aluminum and iron fueled with highly enriched uranium and moderated and reflected by polyethylene were investigated. These critical mass experiments were performed at the Los Alamos Criticality Experiments Facility (LACEF) on the Planet critical assembly. The primary intention of these experiments is to provide supplementary data that can be used to validate and improve criticality data for the Yucca Mountain and the Hanford Storage Waste Tanks Projects. The secondary intention of these experiments is to reduce the H/U ratio and increase the waste material/U ratio from previously published experiments. These experiments were designed to supply data for interlaced waste material/Fuel/Moderator systems on the thermal region. The experiments contained silicon dioxide (SiO{sub 2}), aluminum (Al) and iron (Fe) mixed with 93.23% enriched uranium and moderated and reflected by polyethylene. A base case experiment was also performed with polyethylene-only. This analysis systematically examines uncertainties associated with the critical experiments as they affect the calculated multiplication factor. The systematic analysis is separated into uncertainties due to mass measurements, uncertainties due to fabrication and uncertainties due to composition. Each type of uncertainty is analyzed individually and a total combined uncertainty is derived. The SiO{sub 2}-HEU experiment had a measured k{sub eff} of 0.993, the Al-HEU experiment had a measured k{sub eff} of 0.990, the Fe-HEU had a measured k{sub eff} of 1.000 and the polyethylene-HEU had a measured k{sub eff} of 1.0025. The calculated k{sub eff} values tend to agree well with the experimental values. The sensitivity analysis of these critical experiments yielded a total combined uncertainty on the measured k{sub eff} of {+-}0.0024 for SiO{sub 2}, of {+-}0.0028 for Al, of {+-}0.0026 for Fe, of {+-}0.0020 for polyethylene. (authors)

Loaiza, D.; Sanchez, R. [MS J562, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

2006-07-01T23:59:59.000Z

240

X-ray, K-edge measurement of uranium concentration in reactor fuel plates  

SciTech Connect

Under the Characterization, Monitoring, and Sensor Technology Crosscutting Program, the authors have designed and built a K-edge heavy-metal detector that measures the level of heavy-metal content inside closed containers in a nondestructive, non-invasive way. They have applied this technique to measurement of the amount of uranium in stacks of reactor fuel plates containing nuclear materials of different enrichments and alloys. They have obtained good agreement with expected uranium concentrations ranging from 60 mg/cm{sup 2} to 3,000 mg/cm{sup 2}, and have demonstrated that the instrument can operate in a high radiation field (> 200 mR/hr).

Jensen, T.; Aljundi, T.; Whitmore, C.; Zhong, H.; Gray, J.N.

1997-11-26T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Low enrichment fuel conversion for Iowa State University. Final report  

SciTech Connect

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

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

1996-10-17T23:59:59.000Z

242

CRAD, Radiological Controls- Y-12 Enriched Uranium Operations Oxide Conversion Facility  

Energy.gov (U.S. Department of Energy (DOE))

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Radiation Protection Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

243

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate  

DOE Patents (OSTI)

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Travelli, Armando (Hinsdale, IL)

1988-01-01T23:59:59.000Z

244

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate  

DOE Patents (OSTI)

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Travelli, A.

1985-10-25T23:59:59.000Z

245

Validation of MCNP, a comparison with SCALE: Part 3, Highly enriched uranium oxide systems  

SciTech Connect

This is Part 3 of a series of validation studies dealing with highly enriched uranium systems. For this study only one set of critical experiments involving uranium dioxide have been modeled. Earlier studies address the validation of MCNP for use with highly enriched uranium solutions and metal systems. The calculations of k[sub eff] were performed using MCNP 4. MCNP is a Monte Carlo based transport code which used continuous-energy nuclear data for these calculations. ENDF/B-V cross sections were used for this study. This report also compares the results of MCNP with the results of the CSAS25 module of SCALE 4 using the 27 group ENDF/B-V cross sections. A previous validation study includes information about the CSAS25 module and the resulting data.

Crawford, C.; Palmer, B.M.

1992-10-01T23:59:59.000Z

246

Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Uranium Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit OAS-FS-13-02 October 2012 September 7, 2012 Mr. Gregory Friedman Inspector General U.S. Department of Energy 1000 Independence Avenue, S.W. Room 5D-039 Washington, DC 20585 Dear Mr. Friedman: We have audited the financial statements of the Department of Energy's (the Department) Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) as of and for the year ended September 30, 2011, and have issued our report thereon dated September 7, 2012. In planning and performing our audit of the consolidated financial statements, in accordance with auditing standards generally accepted in the United States of America, we considered the Department's internal control

247

Update on Y-12 national security complex activities to recover enriched uranium in 2007  

SciTech Connect

During Calendar Year 2007, the Y-12 National Security Complex (Y-12) has completed recovery missions that resulted in the return of highly enriched uranium from Canada and several locations within the United States. These missions were performed in support of the National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI) and the Department of Energy (DOE) Central Scrap Management Office for Uranium (U-CSMO). Additionally, Y-12 completed safety basis revisions for the ES-3100 shipping package which resulted in the issuance of a Certificate of Compliance (CoC) from the United States Nuclear Regulatory Commission and a Competent Authority Certificate (CAC) from the United States Department of Transportation for air transport of highly enriched uranium in the form of un- irradiated TRIGA pellets. This certification of the ES-3100 will now allow GTRI to perform recoveries of limited quantities of fresh HEU TRIGA that have been identified at several locations. (author)

Eddy, Becky; Andes, Trent; Dunavant, Randy [Y-12 National Security Complex, Oak Ridge, TN (United States)

2008-07-15T23:59:59.000Z

248

Tritium production using highly enriched fuel  

SciTech Connect

Preliminary studies utilizing the MOFDA code have been made for tritium production at the K reactors using 33 and 35 grams per foot oralloy (93.5% U-235) in aluminum in conjunction with standard K5N and K5E fuel elements, respectively. For this report, it was assumed that all tritium would be produced in discrete charges of LiAl target elements. It is intended that the study will be extended at some later time to include LiAl splines. The analysis includes the effect of coolant loss on reactivity for hot-or-cold and green-or-exposed conditions for several oralloy loading fractions.

Miller, R.L.

1967-12-18T23:59:59.000Z

249

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Albright, D. , Plutonium and highly enriched uranium, 1996 :and swelling in uranium-plutonium mixed nitride fuels.products and to extract plutonium or any other actinide from

Heidet, Florent

2010-01-01T23:59:59.000Z

250

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

SciTech Connect

The US and UK have been separately working on the development of a NDA instrument to determine the enrichment of gaseous UF/sub 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 ..gamma..-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.

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

1985-01-01T23:59:59.000Z

251

US Department of Energy Uranium Enrichment Activity. Financial statements, September 30, 1991 and 1990  

SciTech Connect

KPMG Peat Marwick (KPMG), Certified Public Accountants, has completed its audit of the Department of Energy`s Uranium Enrichment Activity (UEA) financial.statements as of September 30, 1991. The purpose of the audit was to determine whether (1) the financial statements were presented fairly in accordance with applicable accounting principles, (2) the auditee complied with all applicable laws and regulations that may have materially affected the financial statements, and (3) the internal accounting controls, taken as a whole, were adequate. The US Government, through the Department of Energy (DOE) and the management and operating contractor, operates the UEA to enrich uranium hexafluoride in the isotope U-235 for commercial power reactor operators, as further discussed in note 1 of the financial statements. The enrichment of uranium for Government program users, which had been a function of UEA, was transferred outside the UEA affective September 30, 1991, as described in note 3 of the financial statements. UEA is a part of DOE and does not exist as a separate legal entity. For financial reporting purposes, the entity is defined as those activities which provide enriching services to its customers. The financial statements are prepared by extracting and adjusting UEA related data from the financial records of DOE and its contractors.

Not Available

1992-06-16T23:59:59.000Z

252

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.  

Science Conference Proceedings (OSTI)

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 Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertain

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

2010-09-30T23:59:59.000Z

253

Spatially Resolved Strain Fields in Nuclear Fuel Plates Determined ...  

Science Conference Proceedings (OSTI)

Abstract Scope, The U.S. Reduced Enrichment for Research and Test Reactors program converts research reactors which utilize highly enriched uranium fuel to  ...

254

Prompt Neutron Decay for Delayed Critical Bare and Natural-Uranium-Reflected Metal Spheres of Plutonium and Highly Enriched Uranium  

Science Conference Proceedings (OSTI)

Prompt neutron decay at delayed criticality was measured by Oak Ridge National Laboratory for uranium-reflected highly enriched uranium (HEU) and Pu metal spheres (FLATTOP), for an unreflected Pu metal (4.5% {sup 240}Pu) sphere (JEZEBEL) at Los Alamos National Laboratory (LANL) and for an unreflected HEU metal sphere at Oak Ridge Critical Experiments Facility. The average prompt neutron decay constants from hundreds of Rossi-{alpha} and randomly pulsed neutron measurements with {sup 252}Cf at delayed criticality are as follows: 3.8458 {+-} 0.0016 x 10{sup 5} s{sup -1}, 2.2139 {+-} 0.0022 x 10{sup 5} s{sup -1}, 6.3126 {+-} 0.0100 x 10{sup 5} s{sup -1}, and 1.1061 {+-} 0.0009 x 10{sup 6} s{sup -1}, respectively. These values agree with previous measurements by LANL for FLATTOP, JEZEBEL, and GODIVA I as follows: 3.82 {+-} 0.02 x 10{sup 5} s{sup -1} for a uranium core; 2.14 {+-} 0.05 x 10{sup 5} s{sup -1} and 2.29 x 10{sup 5} s{sup -1} (uncertainty not reported) for a plutonium core; 6.4 {+-} 0.1 x 10{sup 5} s{sup -1}, and 1.1 {+-} 0.1 x 10{sup 6} s{sup -1}, respectively, but have smaller uncertainties because of the larger number of measurements. For the FLATTOP and JEZEBEL assemblies, the measurements agree with calculations. Traditionally, the calculated decay constants for the bare uranium metal sphere GODIVA I and the Oak Ridge Uranium Metal Sphere were higher than experimental by {approx}10%. Other energy-dependent quantities for the bare uranium sphere agree within 1%.

Mihalczo, John T [ORNL

2011-01-01T23:59:59.000Z

255

Uranium Enrichment Measurements without Calibration Using Gamma Rays Above 100 keV  

DOE Green Energy (OSTI)

The verification of UF{sub 6} shipping cylinders is an important activity in routine safeguards inspections. Current measurement methods using either sodium-iodide or high-purity germanium detectors require calibrations that are not always appropriate for field measurements, because of changes in geometry or container wall thickness. The introduction of the MGAU code demonstrated the usefulness of intrinsically calibrated measurements for inspections. MGAU uses the 100-keV region of the uranium gamma-ray spectrum. The thick walls of UF{sub 6} shipping cylinders and the low-energy analysis preclude the routine use of MGAU for these measurements. We have developed a uranium enrichment measurement method for measurements using high-purity germanium detectors, which do not require calibration, and uranium gamma rays above 100 keV. The method uses seven gamma rays from {sup 235}U and {sup 238}U to determine their relative detection efficiency intrinsically and with an additional gamma ray from {sup 234}U, the relative abundance of these three uranium isotopes. The method uses a function that describes the basic physical processes that predominantly determine the relative detection efficiency curve. These are the detector efficiency, the absorption by the cylinder wall, and the self-absorption by the uranium contents. We will describe this model and initial testing on various uranium materials and detector types.

Ruhter, W D; Wang, T F; Hayden, C

2001-09-27T23:59:59.000Z

256

Isotope correlation studies relative to high enrichment test reactor fuels  

SciTech Connect

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

Maeck, W.J.; Tromp, R.L.; Duce, F.A.; Emel, W.A.

1978-06-01T23:59:59.000Z

257

Evaluation of oxygen-enrichment system for alternative fuel vehicles  

DOE Green Energy (OSTI)

This report presents results on the reduction in exhaust emissions achieved by using oxygen-enriched intake air on a flexible fuel vehicle (FFV) that used Indolene and M85 as test fuels. The standard federal test procedure (FTP) and the US Environmental Protection Agency`s (EPA`s) off-cycle (REP05) test were followed. The report also provides a review of literature on the oxygen membrane device and design considerations. It presents information on the sources and contributions of cold-phase emissions to the overall exhaust emissions from light-duty vehicles (LDVs) and on the various emission standards and present-day control technologies under consideration. The effects of oxygen-enriched intake air on FTP and off-cycle emissions are discussed on the basis of test results. Conclusions are drawn from the results and discussion, and different approaches for the practical application of this technology in LDVs are recommended.

Poola, R.B.; Sekar, R.R.; Ng, H.K.

1995-12-01T23:59:59.000Z

258

THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE  

DOE Green Energy (OSTI)

Onsager's analysis of the hydrodynamics of fluid circulation in the boundary layer on the rotor wall of a gas centrifuge is reviewed. The description of the flow in the boundary layers on the top and bottom end caps due to Carrier and Maslen is summarized. The method developed by Wood and Morton of coupling the flow models in the rotor wall and end cap boundary layers to complete the hydrodynamic analysis of the centrifuge is presented. Mechanical and thermal methods of driving the internal gas circulation are described. The isotope enrichment which results from the superposition of the elementary separation effect due to the centrifugal field in the gas and its internal circulation is analyzed by the Onsager-Cohen theory. The performance function representing the optimized separative power of a centrifuge as a function of throughput and cut is calculated for several simplified internal flow models. The use of asymmetric ideal cascades to exploit the distinctive features of centrifuge performance functions is illustrated.

Olander, Donald R.

1981-03-01T23:59:59.000Z

259

International safeguards at the feed and withdrawal area of a gas centrifuge uranium enrichment plant  

SciTech Connect

This paper discusses the application of International Atomic Energy Agency (IAEA) safeguards at a model gas centrifuge uranium enrichment plant designed for the production of low-enriched uranium; particular emphasis is placed upon the verification by the IAEA of the facility material balance accounting. After reviewing the IAEA safeguards objectives and concerns at such a plant, the paper describes the material accountancy performed by the facility operator, and discusses strategies by which the operator might attempt to divert a portion of the declared nuclear materials. Finally, the paper discusses the verification of the declared material balance, including sampling strategies, attributes and variables measurements, and nondestructive measurements to improve the efficiency of the inspection measures.

Gordon, D.M.; Sanborn, J.B.

1979-01-01T23:59:59.000Z

260

Reactor fuel conversion assistance request. Technical progress report, August 15, 1992--May 14, 1993  

SciTech Connect

This report is a summary of the progress that has been made on the preparations required to convert the WSU TRIGA reactor from High Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel.

Tripard, G.E.

1993-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Reactor fuel conversion assistance request: Technical progress report, August 15, 1992-December 31, 1994  

SciTech Connect

This report is a summary of the progress that has been made on the preparations required to convert the WSU TRIGA reactor from High Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel.

Tripard, G.E.

1994-12-31T23:59:59.000Z

262

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

Science Conference Proceedings (OSTI)

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 blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including unenriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to diver

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

2008-12-23T23:59:59.000Z

263

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

SciTech Connect

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

G. Youinou; S. Bays

2009-05-01T23:59:59.000Z

264

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  

SciTech Connect

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.

Not Available

1994-01-01T23:59:59.000Z

265

22.351 Systems Analysis of the Nuclear Fuel Cycle, Spring 2003  

E-Print Network (OSTI)

In-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, thorium and other fuel types, ...

Kazimi, Mujid S.

266

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

SciTech Connect

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, UO{sub 3}, was used in the testing. Next, NU made up of UO{sub 3} and DU made up of UO{sub 2} was used in the test work. In every test, the blend achieved was characterized by spatial sampling of the ground product and analyzing for {sup 235}U concentration. The test work proved that these uranium oxide materials can be blended successfully. The spatial concentration was found to be uniform. Next, sintered thorium oxide pellets were used as surrogate for light water breeder reactor pellets (LWBR). To simulate LWBR pellet dispositioning, the thorium oxide pellets were first ground to a powder form and then the powder was blended with NU. In these tests also the concentration of {sup 235}U and {sup 232}Th in blended products fell within established limits proving the success of RM-2 milling technology. RM-2 milling technology is applicable to any dry radioactive waste, especially brittle solids that can be ground up and mixed with the non-radioactive stock.

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

2003-08-01T23:59:59.000Z

267

On the application of IAEA safeguards to plutonium and highly enriched uranium from military inventories  

SciTech Connect

Progress toward the reduction of nuclear arsenals may render surplus hundreds of tonnes of plutonium and highly enriched uranium by the end of the century. None of the acknowledged nuclear weapon states (NWS) is under a specific obligation to submit surplus military inventories to international control. However, inviting the International Atomic Energy Agency (IAEA) to apply safeguards to the plutonium and highly enriched uranium (HEU) released from military use could contribute to building confidence as part of the reductions currently envisaged and could encourage further steps within the states currently planning reductions or by other NWS. If invited, specific arrangements for the application of IAEA safeguards to plutonium and highly enriched uranium from military inventories would be determined by: the institutional provisions adopted; the specified verification requirements; the amounts and forms of plutonium and HEU and the types of facilities to be safeguarded; facility-specific features for the control and accounting of the plutonium and HEU; and the number of facilities where safeguards will be applied. These considerations would be used to establish the most appropriate verificiation arrangements, including the technology to be employed and inspection scheduling arrangements, to provide effective and efficient safeguards. If an invitation is made, the IAEA Board of Governors must approve of the obligations and commitments of the states involved and of the financial arrangements that will ensure the safeguards can be implemented as agreed. 2 tabs.

Shea, T.E. (International Atomic Energy Agency, Wagramerstrasse, Vienna (Austria))

1993-01-01T23:59:59.000Z

268

EA-1123: Transfer of Normal and Low-Enriched Uranium Billets to the United Kingdom, Hanford Site, Richland, Washington  

Energy.gov (U.S. Department of Energy (DOE))

This EA evaluates the environmental impacts of the proposal to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium to the United Kingdom; thus,...

269

Analysis of Enriched Uranium and Weapons Plutonium Reloads for PWRs Using BRACC  

Science Conference Proceedings (OSTI)

Comparisons of the multicycle results demonstrate that the correlation coefficients based on the CASMO3 data were implemented correctly and that the Linear Reactivity Model is acceptably accurate for missed reloads containing both uranium and weapons plutonium fuel. The expanded set of correlation coefficients make BRACC a useful tool for performing multi-cycle in-core fuel management studies of PWR cores containing weapons plutonium.

Alonso, G.; Parish, T.A.

1997-06-05T23:59:59.000Z

270

MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS  

Science Conference Proceedings (OSTI)

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.

FINFROCK SH

2009-12-10T23:59:59.000Z

271

Detailed analysis of uranium silicide dispersion fuel swelling  

SciTech Connect

Swelling of U{sub 3}Si and U{sub 3}Si{sub 2} is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs.

Hofman, G.L.; Ryu, Woo-Seog.

1989-01-01T23:59:59.000Z

272

URANIUM IN ALKALINE ROCKS  

E-Print Network (OSTI)

combine to indicate uranium enrichment of an alkaline magma.uranium, the Ilfmaussaq intrusion contains an unusually high enrichment

Murphy, M.

2011-01-01T23:59:59.000Z

273

Nuclear & Uranium  

U.S. Energy Information Administration (EIA)

Nuclear & Uranium. Uranium fuel ... nuclear reactors, generation, spent fuel. Total Energy. Comprehensive data summaries, comparisons, analysis, and projections ...

274

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents (OSTI)

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Forsberg, C.W.

1998-11-03T23:59:59.000Z

275

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents (OSTI)

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotonically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Forsberg, Charles W.

1997-12-01T23:59:59.000Z

276

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents (OSTI)

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Forsberg, Charles W. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

277

Microstructure Characterization and Processing of U-Mo Alloy Fuels ...  

Science Conference Proceedings (OSTI)

molybdenum (Mo) fuels have been identified as a potential replacement for highly enriched uranium (HEU) dispersion fuels in high performance research ...

278

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

SciTech Connect

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.

Knight, J.R.

1984-11-01T23:59:59.000Z

279

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

SciTech Connect

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.

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

2009-01-01T23:59:59.000Z

280

Nuclear & Uranium - U.S. Energy Information Administration (EIA)  

U.S. Energy Information Administration (EIA)

The U.S. relies on foreign uranium, enrichment services to fuel its nuclear power plants August 28, 2013. See all new releases in EIA ...

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

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

Science Conference Proceedings (OSTI)

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.

NONE

1995-05-01T23:59:59.000Z

282

Criticality safety considerations for MSRE fuel drain tank uranium aggregation  

SciTech Connect

This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations. Since the salt was transferred to the FDTs in 1969, radiological and chemical reactions have been converting the uranium and fluorine in the salt to UF{sub 6} and free fluorine. Significant amounts of uranium (at least 3 kg) and fluorine have migrated out of the FDTs and into the off-gas system (OGS) and the auxiliary charcoal bed (ACB). The loss of uranium and fluorine from the salt changes the chemical properties of the salt sufficiently to possibly allow the reduction of the UF{sub 4} in the salt to uranium metal as the salt is remelted prior to removal. It has been postulated that up to 9 kg of the maximum 19.4 kg of uranium in one FDT could be reduced to metal and concentrated. This study shows that criticality becomes a concern when more than 5 kg of uranium concentrates to over 8 wt% of the salt in a favorable geometry.

Hollenbach, D.F.; Hopper, C.M. [Oak Ridge National Lab., TN (United States). Computational Physics and Engineering Div.

1997-03-01T23:59:59.000Z

283

Oxidation Protection of Uranium Nitride Fuel using Liquid Phase Sintering  

SciTech Connect

Two methods are proposed to increase the oxidation resistance of uranium nitride (UN) nuclear fuel. These paths are: (1) Addition of USi{sub x} (e.g. U3Si2) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with various compounds (followed by densification via Spark Plasma Sintering or Liquid Phase Sintering) that will greatly increase oxidation resistance. The advantages (high thermal conductivity, very high melting point, and high density) of nitride fuel have long been recognized. The sodium cooled BR-10 reactor in Russia operated for 18 years on uranium nitride fuel (UN was used as the driver fuel for two core loads). However, the potential advantages (large power up-grade, increased cycle lengths, possible high burn-ups) as a Light Water Reactor (LWR) fuel are offset by uranium nitride's extremely low oxidation resistance (UN powders oxidize in air and UN pellets decompose in hot water). Innovative research is proposed to solve this problem and thereby provide an accident tolerant LWR fuel that would resist water leaks and high temperature steam oxidation/spalling during an accident. It is proposed that we investigate two methods to increase the oxidation resistance of UN: (1) Addition of USi{sub x} (e.g. U{sub 3}Si{sub 2}) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with compounds (followed by densification via Spark Plasma Sintering) that will greatly increase oxidation resistance.

Dr. Paul A. Lessing

2012-03-01T23:59:59.000Z

284

22.251 / 22.351 Systems Analysis of the Nuclear Fuel Cycle, Fall 2005  

E-Print Network (OSTI)

This course provides an in-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, ...

Kazimi, Mujid S.

285

Fuels I - Programmaster.org  

Science Conference Proceedings (OSTI)

Mar 5, 2013 ... Homogenization of Low Enriched Uranium-10 wt. pct Molybdenum Alloy Monolithic Fuel Foils: Amy Clarke1; Kester Clarke1; David Alexander1; ...

286

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

SciTech Connect

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)

1975-01-01T23:59:59.000Z

287

Process for producing enriched uranium having a {sup 235}U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage  

DOE Patents (OSTI)

An uranium enrichment process capable of producing an enriched uranium, having a {sup 235}U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower {sup 235}U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF{sub 6} tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a {sup 235} U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % {sup 235} U; fluorinating this enriched metallic uranium isotopic mixture to form UF{sub 6}; processing the resultant isotopic mixture of UF{sub 6} in a gaseous diffusion process to produce a final enriched uranium product having a {sup 235}U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low {sup 235}U content UF{sub 6} having a {sup 235}U content of about 0.71 wt. % of the total uranium content of the low {sup 235}U content UF{sub 6}; and converting this low {sup 235}U content UF{sub 6} to metallic uranium for recycle to the atomic vapor laser isotope separation process. 4 figs.

Horton, J.A.; Hayden, H.W. Jr.

1995-05-30T23:59:59.000Z

288

Process for producing enriched uranium having a .sup.235 U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage  

DOE Patents (OSTI)

An uranium enrichment process capable of producing an enriched uranium, having a .sup.235 U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower .sup.235 U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF.sub.6 tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a .sup.235 U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % .sup.235 U; fluorinating this enriched metallic uranium isotopic mixture to form UF.sub.6 ; processing the resultant isotopic mixture of UF.sub.6 in a gaseous diffusion process to produce a final enriched uranium product having a .sup.235 U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low .sup.235 U content UF.sub.6 having a .sup.235 U content of about 0.71 wt. % of the total uranium content of the low .sup.235 U content UF.sub.6 ; and converting this low .sup.235 U content UF.sub.6 to metallic uranium for recycle to the atomic vapor laser isotope separation process.

Horton, James A. (Livermore, CA); Hayden, Jr., Howard W. (Oakridge, TN)

1995-01-01T23:59:59.000Z

289

Office of Environmental Management uranium enrichment decontamination and decommissioning fund financial statements. September 30, 1994 and 1993  

SciTech Connect

The Energy Policy Act of 1992 (Act) transferred the uranium enrichment enterprise to the United States Enrichment Corporation as of July 1, 1993. However, the 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 in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio (diffusion facilities). The Act 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.

Marwick, P.

1994-12-15T23:59:59.000Z

290

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

SciTech Connect

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.

1996-02-21T23:59:59.000Z

291

Parametric Study of Front-End Nuclear Fuel Cycle Costs Using Reprocessed Uranium  

Science Conference Proceedings (OSTI)

This study evaluates front-end nuclear fuel cycle costs assuming that uranium recovered during the reprocessing of commercial light-water reactor (LWR) spent nuclear fuel is available to be recycled and used in the place of natural uranium. This report explores the relationship between the costs associated with using a natural uranium fuel cycle, in which reprocessed uranium (RepU) is not recycled, with those associated with using RepU.

2010-01-26T23:59:59.000Z

292

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

SciTech Connect

This report discusses each of the components that affect the unit cost for enriched uranium; that is, ore costs, U/sub 3/O/sub 8/ to UF/sub 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 /sup 235/U) and for low-enrichment uranium (3.0, 3.2, and 3.5 percent /sup 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.

Thomas, W.E.

1976-04-01T23:59:59.000Z

293

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input  

Science Conference Proceedings (OSTI)

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi [Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134 (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Reserach of Laboratory for Nuclear Reactors, Tokyo Institute of Technology O-okayama, Meguro-ku, Tokyo 152 (Japan)

2012-06-06T23:59:59.000Z

294

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

Science Conference Proceedings (OSTI)

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.

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

2008-09-01T23:59:59.000Z

295

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

SciTech Connect

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

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

2002-02-26T23:59:59.000Z

296

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

SciTech Connect

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.

JOE,J.

2007-07-08T23:59:59.000Z

297

Signatures and Methods for the Automated Nondestructive Assay of UF6 Cylinders at Uranium Enrichment Plants  

Science Conference Proceedings (OSTI)

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These measurements are time-consuming, expensive, and assay only a small fraction of the total cylinder volume. An automated nondestructive assay system capable of providing enrichment measurements over the full volume of the cylinder could improve upon current verification practices in terms of manpower and assay accuracy. Such a station would use sensors that can be operated in an unattended mode at an industrial facility: medium-resolution scintillators for gamma-ray spectroscopy (e.g., NaI(Tl)) and moderated He-3 neutron detectors. This sensor combination allows the exploitation of additional, more-penetrating signatures beyond the traditional 185-keV emission from U-235: neutrons produced from F-19(?,n) reactions (spawned primarily from U 234 alpha emission) and high-energy gamma rays (extending up to 8 MeV) induced by neutrons interacting in the steel cylinder. This paper describes a study of these non-traditional signatures for the purposes of cylinder enrichment verification. The signatures and the radiation sensors designed to collect them are described, as are proof-of-principle cylinder measurements and analyses. Key sources of systematic uncertainty in the non-traditional signatures are discussed, and the potential benefits of utilizing these non-traditional signatures, in concert with an automated form of the traditional 185-keV-based assay, are discussed.

Smith, Leon E.; Mace, Emily K.; Misner, Alex C.; Shaver, Mark W.

2010-08-08T23:59:59.000Z

298

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01T23:59:59.000Z

299

Fuel Rod Cooling in Natural Uranium Reactors  

SciTech Connect

An analysis is presented of the transfer of heat from a cylindrical fuel rod surrounded by a fast flowing coolant in an annular duct, with maximum power output limited by fuel rod temperatures, coolant pressure drop and pumping power requirements. A method is also presented for comparing and evaluating various liquid and gaseous coolants within these limitations. The report also shows and discusses some calculated results obtained for the systems considred in the study of natural U reactors for the production of Pu and useful power (NAA-SR-137).

Trilling, C.A.

1952-01-28T23:59:59.000Z

300

Measurement of highly enriched uranium metal buttons with the high-level neutron coincidence counter operating in the active mode  

SciTech Connect

The portable High-Level Neutron Coincidence Counter is used in the active mode with the addition of AmLi neutron sources to assay the /sup 235/U content of highly enriched metal pieces or buttons. It is concluded that the portable instrument is a practical instrument for assaying uranium metal buttons with masses in the range 1.5 to 4 kg.

Foley, J.E.

1980-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Uranium chloride extraction of transuranium elements from LWR fuel  

DOE Patents (OSTI)

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

1992-08-25T23:59:59.000Z

302

Uranium chloride extraction of transuranium elements from LWR fuel  

DOE Patents (OSTI)

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Miller, William E. (Naperville, IL); Ackerman, John P. (Downers Grove, IL); Battles, James E. (Oak Forest, IL); Johnson, Terry R. (Wheaton, IL); Pierce, R. Dean (Naperville, IL)

1992-01-01T23:59:59.000Z

303

Uranium chloride extraction of transuranium elements from LWR fuel  

Science Conference Proceedings (OSTI)

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800{degrees}C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

1991-12-31T23:59:59.000Z

304

TEM Characterization of Irradiated RERTR Dispersion Fuels  

Science Conference Proceedings (OSTI)

Abstract Scope, The USA fuels program on Reduced Enrichment Research and Test Reactors ... Segregation of Ru to Edge Dislocations in Uranium Dioxide.

305

Low enrichment fuel conversion for Iowa State University  

SciTech Connect

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

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

1990-08-01T23:59:59.000Z

306

Criticality issues with highly enriched fuels in a repository environment  

SciTech Connect

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

Taylor, L.L. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Sanchez, L.C.; Rath, J.S. [Sandia National Labs., Albuquerque, NM (United States)

1998-03-01T23:59:59.000Z

307

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels  

SciTech Connect

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Ackerman, John P. (Downers Grove, IL); Miller, William E. (Naperville, IL)

1989-01-01T23:59:59.000Z

308

Developing fuel management capabilities based on coupled Monte Carlo depletion in support of the MIT Research Reactor (MITR) conversion  

E-Print Network (OSTI)

Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to ...

Romano, Paul K. (Paul Kollath)

2009-01-01T23:59:59.000Z

309

Improved accountability method for measuring enriched uranium in H-Canyon dissolver solution at the Savannah River Site  

SciTech Connect

At the Savannah River Site (SRS), accountability measurement of enriched uranium dissolved in H-Canyon is performed using isotope dilution mass spectrometry (IDMS). In the IDMS analytical method, a known quantity of uranium{sup 233} is added to the sample solution containing enriched uranium and fission products. The resulting uranium mixture must first be purified using a separation technique in the shielded analytical(``hot``) cells to lower radioactivity levels by removing fission products. Following this purification, the sample is analyzed by mass spectrometry to determine the total uranium content and isotopic abundance. The magnitude of the response of each uranium isotope in the sample solution and the response of the U{sup 233} spike is measured. By ratioing these responses, relative to the known quantity of the U{sup 233} spike, the uranium content can be determined. A hexane solvent extraction technique, used for years at SRS to remove fission products prior to the mass spectrometry analysis of uranium, has several problems. The hexone method is tedious, requires additional sample clean-up after the purified sample is removed from the shielded cells and requires the use of Resource Conservation and Recovery Act (RCRA)-listed hazardous materials (hexone and chromium compounds). A new high speed separation method that enables a rapid removal of fission products in a shielded cells environment has been developed by the SRS Central Laboratory to replace the hexone method. The new high speed column extraction chromatography technique employs applied vacuum and columns containing tri (2-ethyl-hexyl) phosphate (TEHP) solvent coated on a small particle inert support (SM-7 Bio Beads). The new separation is rapid, user friendly, eliminates the use of the RCA-listed hazardous chemicals and reduces the amount of solid waste generated by the separation method. 2 tabs. 4 figs.

Maxwell, S.L. III; Satkowski, J.; Mahannah, R.N.

1992-08-01T23:59:59.000Z

310

Improved accountability method for measuring enriched uranium in H-Canyon dissolver solution at the Savannah River Site  

SciTech Connect

At the Savannah River Site (SRS), accountability measurement of enriched uranium dissolved in H-Canyon is performed using isotope dilution mass spectrometry (IDMS). In the IDMS analytical method, a known quantity of uranium{sup 233} is added to the sample solution containing enriched uranium and fission products. The resulting uranium mixture must first be purified using a separation technique in the shielded analytical( hot'') cells to lower radioactivity levels by removing fission products. Following this purification, the sample is analyzed by mass spectrometry to determine the total uranium content and isotopic abundance. The magnitude of the response of each uranium isotope in the sample solution and the response of the U{sup 233} spike is measured. By ratioing these responses, relative to the known quantity of the U{sup 233} spike, the uranium content can be determined. A hexane solvent extraction technique, used for years at SRS to remove fission products prior to the mass spectrometry analysis of uranium, has several problems. The hexone method is tedious, requires additional sample clean-up after the purified sample is removed from the shielded cells and requires the use of Resource Conservation and Recovery Act (RCRA)-listed hazardous materials (hexone and chromium compounds). A new high speed separation method that enables a rapid removal of fission products in a shielded cells environment has been developed by the SRS Central Laboratory to replace the hexone method. The new high speed column extraction chromatography technique employs applied vacuum and columns containing tri (2-ethyl-hexyl) phosphate (TEHP) solvent coated on a small particle inert support (SM-7 Bio Beads). The new separation is rapid, user friendly, eliminates the use of the RCA-listed hazardous chemicals and reduces the amount of solid waste generated by the separation method. 2 tabs. 4 figs.

Maxwell, S.L. III; Satkowski, J.; Mahannah, R.N.

1992-01-01T23:59:59.000Z

311

Depleted Uranium  

NLE Websites -- All DOE Office Websites (Extended Search)

Depleted Uranium Depleted Uranium Depleted Uranium line line Uranium Enrichment Depleted Uranium Health Effects Depleted Uranium Depleted uranium is uranium that has had some of its U-235 content removed. Over the last four decades, large quantities of uranium were processed by gaseous diffusion to produce uranium having a higher concentration of uranium-235 than the 0.72% that occurs naturally (called "enriched" uranium) for use in U.S. national defense and civilian applications. "Depleted" uranium is also a product of the enrichment process. However, depleted uranium has been stripped of some of its natural uranium-235 content. Most of the Department of Energy's (DOE) depleted uranium inventory contains between 0.2 to 0.4 weight-percent uranium-235, well

312

Uranium industry annual 1998  

SciTech Connect

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.

NONE

1999-04-22T23:59:59.000Z

313

A PROCESS FOR THE RECOVERY OF URANIUM FROM NUCLEAR FUEL ELEMENTS USING FLUID-BED DRYING AND VOLATILITY TECHNIQUES  

SciTech Connect

A process scheme for the recovery of uranium from fuel elements has been developed. The scheme combines continuous fluid-bed drying and fluoride volatility techniques after initial dissolution of the fuel element in the appropriate aqueous system, hence the designation ADF, Aqueous Dry Fluorination Process. The application of this process to the recovery of uranium from highly enriched, low uranium-zirconium alloy plate-type fuels is described. ln the process, the feed solution is sprayed horizontally through a two-fluid nozzle and is atomized directly in the heated fluidized bed. The spray droplets are dried on the fluidized particles and form a dense coating. Excessive particle growth was limited by the use of air attrition-jets inserted directly in the bed. Aqueous hydrofluoric acid solutions containing l.2 to 3.6 M zirconiuni, 0.007 to 0.03 M uranium, and free acid concentrations from 1 to about l0 M were successfully processed in a 6-in.-diameter Inconel fluid-bed spray dryer. Rates equivalent to about 3.l kg/hr of zirconium were achieved, 160 ml/min with the most concentrated feed solution. Experiments were successfully carried out from 240 to 450 deg C. A new design for a two-fluid nozzle was developed. Extensive work was done to identify the various zirconium fluoride compounds formed. The granular dryer product was subsequently fluorinated at temperatures to 600 deg C in fluid beds and to 700 deg C in static beds to remove the uranium as the volatile hexafluoride. About 90 to 95% uranium removal was consistently achieved near 600 deg C. The relatively low uranium recovery under these conditions is a disadvantage for the application to zirconium-base fuels. It was found necessary to resort to static beds and higher temperatures to achieve greater removal. Since the fluorine attack on nickel, the material of construction, is prohibitive at temperatures above 600 deg C, a disposable fluorinator concept for use with static beds is described. Results of corrosion studies are reported. A preliminary chemical flowsheet with a design capacity of 1l00 kg of uranium (93% enriched) annually is presented. (auth)

Levitz, N.; Barghusen, J.; Carls, E.; Jonke, A.A.

1961-11-01T23:59:59.000Z

314

Highly Enriched Uranium Metal Annuli and Cylinders with Polyethylene Reflectors and/or Internal Polyethylene Moderator  

SciTech Connect

A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, experiments of uranium metal annuli with and without polyethylene reflectors and with the central void region either empty or filled with polyethylene were evaluated under ICSBEP Identifier HEU-MET-FAST-076. The outer diameter of the uranium annuli varied from 9 to 15 inches in two-inch increments. In addition, there were uranium metal cylinders with diameters varying from 7 to 15 inches with complete reflection and reflection on one flat surface to simulate floor reflection. Most of the experiments were performed between February 1964 and April 1964. Five partially reflected (reflected on the top only) experiments were assembled in November 1967, but are judged by the evaluators not to be of benchmark quality. Twenty-four of the twenty-five experiments have been determined to have fast spectra. The only exception has a mixed spectrum. Analyses were performed in which uncertainty associated with five different parameters associated with the uranium parts and three associated with the polyethylene parts was evaluated. Included were uranium mass, height, diameter, isotopic content, and impurity content and polyethylene mass, diameter, and impurity content. There were additional uncertainties associated with assembly alignment, support structure, and the value for ßeff. In addition to the idealizations made by the experimenters (removal of a diaphragm), a few simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. Simplifications included omission of the support structure, possible surrounding equipment, and the walls, floor, and ceiling of the experimental cell. Bias values that result from these simplifications were determined and associated uncertainty in the bias values were included in the overall uncertainty in benchmark keff values. Bias values ranged from 0.0002 ?k to 0.0093 ?k below the experimental value. Overall uncertainties range from ? 0.0002 to ? 0.0011. Major contributors to the overall uncertainty include uncertainty in the support structure and the polyethylene parts. A comparison of experimental, benchmark-model, and MCNP-model keff values is shown in Figure 1. The experimental keff values are derived from the original reactivities reported by the principal experimentalist. The benchmark-model keff values are the experimental keff values adjusted to account for biases that were introduced by removing the support structure and surroundings. The MCNP-model keff values are simply the values found from MCNP calculations using the benchmark specifications and ENDF/B-VI cross-section data. Figure 1. Comparison of Experimental, Benchmark-Model and MCNP-Model keff value. Calculated results for most of the experiments are

Tyler Sumner; J. Blair Briggs; Leland Montierth

2007-05-01T23:59:59.000Z

315

Validation of KENO V.a for highly enriched uranium systems with hydrogen and/or carbon moderation  

SciTech Connect

This paper describes the validation in accordance with ANSI/ANS-8.1-1983(R1988) of KENO V.a using the 27-group ENDF/B-IV cross-section library for systems containing highly-enriched uranium, carbon, and hydrogen and for systems containing highly-enriched uranium and carbon with high carbon to uranium (C/U) atomic ratios. The validation has been performed for two separate computational platforms: an IBM 3090 mainframe and an HP 9000 Model 730 workstation, both using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software (NCSS) code package. Critical experiments performed at the Oak Ridge Critical Experiments Facility, in support of the Rover reactor program, and at the Pajarito site at Los Alamos National Laboratory were 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. Calculated values of k{sub eff} for the Rover experiments, which contain uranium, carbon, and hydrogen, are between 1.0012 {+-} 0.0026 and 1.0245 {+-} 0.0023. Calculation of the Los Alamos experiments, which contain uranium and carbon at high C/U ratios, yields values of k{sub eff} between 0.9746 {+-} 0.0028 and 0.9983 {+-} 0.0027. Safety criteria can be established using this data for both types of systems.

Elliott, E.P.; Vornehm, R.G. [Oak Ridge Y-12 Plant, TN (United States); Dodds, H.L. Jr. [Univ. of Tennessee, Knoxville, TN (United States). Nuclear Engineering Dept.

1993-06-04T23:59:59.000Z

316

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  

SciTech Connect

A new section, 10 CFR 74.33, has been added to the material control and accounting (MC 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 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.

Moran, B.W.; Belew, W.L. (Oak Ridge K-25 Site, TN (United States)); Hammond, G.A.; Brenner, L.M. (21st Century Industries, Inc., Gaithersburg, MD (United States))

1991-11-01T23:59:59.000Z

317

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

Science Conference Proceedings (OSTI)

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

Not Available

1980-01-01T23:59:59.000Z

318

Microsoft Word - nuclear_fuel_yacout.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

in diameter and 1 cm long. Manufacturing of this form of uranium fuel starts with mined uranium that passes through processes of conversion and enrichment before it is made into...

319

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

SciTech Connect

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.

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

1996-05-01T23:59:59.000Z

320

Physical inventory verification exercise for a highly enriched uranium fabrication facility  

SciTech Connect

The International Atomic Energy Agency, in collaboration with the US Support Program (POTAS), has developed and conducted a training exercise simulating a physical inventory verification (PIV) at a highly enriched uranium (HEU) fabrication facility. This exercise is part of a series sponsored by the POTAS program, including PIVs at light-water reactors and plutonium fabrication facilities. The first HEU exercise took place in September 1985 at Los Alamos National Laboratory and a second is scheduled for Spring, 1987 at JRC, ISPRA. The main objectives of these exercises are: to provide the opportunity for inspectors to test and evaluate the use of nondestructive assay (NDA) equipment and computer software under conditions similar to those found during actual inspections; to use the data generated to evaluate different inspection procedures and strategies; and to exchange ideas on PIV procedures between the three operations divisions. Because the exercises are conducted in a neutral environment, free of the time pressure often found in actual inspections, it is possible for the inspectors to achieve the course objectives.

Abedin-Zadeh, R.; Augustson, R.H.

1986-01-01T23:59:59.000Z

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321

Modeling of Fission Neutrons as a Signature for Detection of Highly Enriched Uranium  

SciTech Connect

We present the results of modeling intended to evaluate the feasibility of using neutrons from induced fission in highly enriched uranium (HEU) as a means of detecting clandestine HEU, even when it is embedded in absorbing surroundings, such as commercial cargo. We characterized radiation from induced fission in HEU, which consisted of delayed neutrons at all energies and prompt neutrons at energies above a threshold. We found that for the candidate detector and for the conditions we considered, a distinctive HEU signature should be detectable, given sufficient detector size, and should be robust over a range of cargo content. In the modeled scenario, an intense neutron source was used to induce fissions in a spherical shell of HEU. To absorb, scatter, and moderate the neutrons, we place one layer of simulated cargo between the source and target and an identical layer between the target and detector. The resulting neutrons and gamma rays are resolved in both time and energy to reveal the portion arising from fission. We predicted the dominant reaction rates within calcium fluoride and liquid organic scintillators. Finally, we assessed the relative effectiveness of two common neutron source energies.

Wolford, J K; Frank, M I; Descalle, M

2004-03-09T23:59:59.000Z

322

ES-3100: A New Generation Shipping Container for Bulk Highly Enriched Uranium and Other Fissile Materials  

SciTech Connect

The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping bulk quantities of surplus fissile materials, primarily highly enriched uranium (HEU), over the next 15 to 20 years for disposition purposes. The U.S. Department of Transportation (DOT) specification 6M container is the package of choice for most of these shipments. However, the 6M does not conform to the Type B packaging requirements in the ''Code of Federal Regulations'' (10CFR71) and, for that reason, is being phased out for use in the secure transportation system of DOE. BWXT Y-12 is currently developing a package to replace the DOT 6M container for HEU disposition shipping campaigns. The new package is based on state-of-the-art, proven, and patented insulation technologies that have been successfully applied in the design of other packages. The new package, designated the ES-3100, will have a 50% greater capacity for HEU than the 6M and will be easier to use. Engineering analysis on the new package includes detailed dynamic impact finite element analysis (FEA). This analysis gives the ES-3100 a high probability of complying with regulatory requirements.

Arbital, J.G.; Byington, G.A.; Tousley, D.R.

2004-07-01T23:59:59.000Z

323

Stationary and protable instruments for assay of HEU (highly enriched uranium) solids holdup  

SciTech Connect

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/sub 3/, UO/sub 2/, and UF/sub 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 /sup 235/U gamma ray is the assay signature, and the 60-keV gamma ray from an /sup 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.

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

1987-01-01T23:59:59.000Z

324

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

SciTech Connect

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.

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

2010-04-01T23:59:59.000Z

325

Safeguards by design - industry engagement for new uranium enrichment facilities in the United States  

Science Conference Proceedings (OSTI)

The United States Department of Energy's (DOE's) Office of Nonproliferation and International Security (NA-24) has initiated a Safeguards by Design (SBD) effort to encourage the incorporation of international (IAEA) safeguards features early in the design phase of a new nuclear facility in order to avoid the need to redesign or retrofit the facility at a later date. The main goals of Safeguards by Design are to (1) make the implementation of international safeguards at new civil nuclear facilities more effective and efficient, (2) avoid costly and time-consuming re-design work or retrofits at such facilities and (3) design such facilities in a way that makes proliferation as technically difficult, as time-consuming, and as detectable as possible. The U.S. Nuclear Regulatory Commission (NRC) has recently hosted efforts to facilitate the use of Safeguards by Design for new uranium enrichment facilities currently being planned for construction in the U.S. While SBD is not a NRC requirement, the NRC is aiding the implementation of SBD by coordinating discussions between DOE's NA-24 and industry's facility design teams. More specifically, during their normal course of licensing discussions the NRC has offered industry the opportunity to engage with NA-24 regarding SBD.

Demuth, Scott F [Los Alamos National Laboratory; Grice, Thomas [NRC; Lockwood, Dunbar [DOE/NA-243

2010-01-01T23:59:59.000Z

326

Bugs boost Cold War clean-up: Bacteria could scrub uranium from sites contaminated decades ago. updated at midnight GMTtoday is friday, november 14  

E-Print Network (OSTI)

.7% (no enrichment) and around 20% · Large amount of depleted uranium results from enrichment Energy an integrated facility (Integral Fast Reactor), where only small amounts of natural uranium or waste depleted nuclear fuel cycles Ore · All fuel cycles begin with uranium and/or thorium which are the only naturally

Lovley, Derek

327

DOE/EIS-0240-SA-1: Supplement Analysis for the Disposition of Surplus Highly Enriched Uranium (October 2007)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

0-SA1 0-SA1 SUPPLEMENT ANALYSIS DISPOSITION OF SURPLUS HIGHLY ENRICHED URANIUM October 2007 U.S. Department of Energy National Nuclear Security Administration Office of Fissile Materials Disposition Washington, D.C. i TABLE OF CONTENTS 1.0 Introduction and Purpose .................................................................................................................1 2.0 Background......................................................................................................................................1 2.1 Scope of the HEU EIS............................................................................................................ 2 2.2 Status of Surplus HEU Disposition Activities .......................................................................

328

Metallography of pitted aluminum-clad, depleted uranium fuel  

Science Conference Proceedings (OSTI)

The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact.

Nelson, D.Z.; Howell, J.P.

1994-12-01T23:59:59.000Z

329

Converting {sup 99}Mo production from high- to low-enriched uranium  

SciTech Connect

This paper discusses efforts towards LEU substitution in two HEU targets. One type is the Cintichem target, a closed cylinder with a thin coating of uranium dioxide electroplated ion the inside wall. To successfully increase the amount of uranium per target, we are developing a target that uses LEU metal foil. Uranium surface preparation is discussed.

Vandegrift, G.F.; Conner, C.J.; Sedlet, J.; Wygmans, D.G.

1997-09-01T23:59:59.000Z

330

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

SciTech Connect

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

Adams, T.M.

1998-11-25T23:59:59.000Z

331

Licensed fuel facility status report: Inventory difference data, January 1986-June 1986  

SciTech Connect

NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233.

1987-02-01T23:59:59.000Z

332

Licensed fuel facility status report: Inventory difference data, July 1986-December 1986  

SciTech Connect

NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233.

1987-08-01T23:59:59.000Z

333

Synthesis of Uranium Trichloride for the Pyrometallurgical Processing of Used Nuclear Fuel  

SciTech Connect

The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results and conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.

B.R. Westphal; J.C. Price; R.D. Mariani

2011-11-01T23:59:59.000Z

334

Comparison of CdTe and CdZnTe Detectors for Field Determination of Uranium Isotopic Enrichments  

DOE Green Energy (OSTI)

A performance comparison of a CdTe and a CdZnTe detector when exposed to uranium samples of various isotopic enrichments has been performed. These high-resolution detectors can assist in the rapid determination of uranium isotopic content of illicit material. Spectra were recorded from these room temperature semiconductor detectors with a portable multi-channel analyzer, both in the laboratory and in a field environment. Both detectors were operated below ambient temperature using the vendor supplied thermoelectric coolers. Both detectors had nominally the same active volume (18 mm3 for the CdZnTe and 25 mm3 for the CdTe detector) and resolution. Spectra of samples of known isotopic content were recorded at fixed geometries. An evaluation of potential signature g rays for the detection of enriched uranium was completed. Operational advantages and disadvantages of each detector are discussed. There is a need to improve the detection sensitivity during the interdiction of special nuclear materials (SNM) for increased homeland protection. It is essential to provide additional tools to first responders and law enforcement personnel for assessing nuclear and radiological threats.

Hofstetter, KJ

2004-01-23T23:59:59.000Z

335

A MULTI-COUNTRY ANALYSIS OF LIFECYCLE EMISSIONS FROM TRANSPORTATION FUELS AND MOTOR VEHICLES  

E-Print Network (OSTI)

Association, “Uranium Enrichment,” www.world- nuclear.org/0.30 for military high-enriched uranium. Uranium enrichment.international trade in uranium enrichment services (measured

Delucchi, Mark

2005-01-01T23:59:59.000Z

336

A Multi-Country Analysis of Lifecycle Emissions From Transportation Fuels and Motor Vehicles  

E-Print Network (OSTI)

Association, “Uranium Enrichment,” www.world- nuclear.org/0.30 for military high-enriched uranium. Uranium enrichment.international trade in uranium enrichment services (measured

Delucchi, Mark

2005-01-01T23:59:59.000Z

337

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

SciTech Connect

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.

Rothe, R.E.

1997-12-01T23:59:59.000Z

338

Uranium industry annual 1995  

SciTech Connect

The Uranium Industry Annual 1995 (UIA 1995) provides current statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1995 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 period 1986 through 2005 as collected on the Form EIA-858, ``Uranium Industry Annual Survey``. Data collected on the ``Uranium Industry Annual Survey`` provide a comprehensive statistical characterization of the industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1995, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. Data on uranium raw materials activities for 1986 through 1995 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 2005, 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. The methodology used in the 1995 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. For the reader`s convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix D along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 14 figs., 56 tabs.

NONE

1996-05-01T23:59:59.000Z

339

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents (OSTI)

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Pruett, D.J.; McTaggart, D.R.

1983-08-31T23:59:59.000Z

340

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents (OSTI)

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc.sup.+7 therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Pruett, David J. (Knoxville, TN); McTaggart, Donald R. (Knoxville, TN)

1984-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

CORRELATION OF CRITICAL MASS DATA ON LIGHT WATER MODERATED, FULLY ENRICHED URANIUM, STAINLESS STEEL REACTORS. PART I  

SciTech Connect

Experimental data were collected on over 70 light water moderated, fully enriched uranium, stainless steel, critical cores. An equation for the critical mass of cores with a buckling of 0.007 cm/sup -2/ that is lineally dependent on stainless steel volume fraction and grams of B/sup 10/ was compared with available critical experiments and found to yield reasonable results. A correlation method, relating buckling to ( xi SIGMA /sub s// SIGMA /sub a/) was found to fit the available experiments. (auth)

Lee, D.H.

1962-07-17T23:59:59.000Z

342

EA-1255: Project Partnership Transportation of Foreign-Owned Enriched  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5: Project Partnership Transportation of Foreign-Owned 5: Project Partnership Transportation of Foreign-Owned Enriched Uranium from the Republic of Georgia EA-1255: Project Partnership Transportation of Foreign-Owned Enriched Uranium from the Republic of Georgia SUMMARY This EA evaluates the environmental impacts for the proposal to transport 5.26 kilograms of enriched uranium-23 5 in the form of nuclear fuel, from the Republic of Georgia to the United Kingdom. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 30, 1998 EA-1255: Finding of No Significant Impact Project Partnership Transportation of Foreign-Owned Enriched Uranium from the Republic of Georgia April 30, 1998 EA- 1255: Finding of No Significant Impact Project Partnership Transportation of Foreign-Owned Enriched Uranium from

343

Two U.S. University Research Reactors to be Converted From Highly Enriched  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

U.S. University Research Reactors to be Converted From Highly U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium April 11, 2005 - 11:34am Addthis WASHINGTON, D.C. - As part of the Bush administration's aggressive effort to reduce the amount of weapons-grade nuclear material worldwide, Secretary of Energy Samuel W. Bodman announced today that the Department of Energy (DOE) has begun to convert research reactors from using highly-enriched uranium (HEU) to low-enriched uranium fuel (LEU) at the University of Florida and Texas A&M University. This effort, by DOE's National Nuclear Security Administration (NNSA) and the Office of Nuclear Energy, Science and Technology, are the latest steps

344

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents (OSTI)

A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Ackerman, John P. (Downers Grove, IL)

1992-01-01T23:59:59.000Z

345

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents (OSTI)

This report discusses a process for separating uranium values and transuranic values from fission products containing rare earth values when the values which are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is re-established.

Ackerman, J.P.

1991-01-01T23:59:59.000Z

346

Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium Oxide  

SciTech Connect

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) 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 certified reference material (CRM) 149 [Uranium (93% Enriched) Oxide-U{sub 3}O{sub 8} Standard for Neutron Counting Measurements], a unit of which was delivered to LLNL in October of 1999. This paper describes the approach to calibration of the LLNL passive-active neutron drum (PAN) shuffler for measurement of poorly measured/unmeasured HEU oxide inventory. Included are discussions of (1) the calibration effort, including the development of the mass calibration curve; (2) the results from an axial and radial mapping of the detector response over a wide region of the PAN shuffler counting chamber, and (3) an error model for the total (systematic + random) uncertainty in the predicted mass that includes the uncertainties in calibration and sample position.

Mount, M.; Glosup, J.; Cochran, C.; Dearborn, D.; Endres, E.

2000-06-16T23:59:59.000Z

347

SEGREGATION IN URANIUM-ALUMINUM ALLOYS AND ITS EFFECT ON THE FUEL LOADING OF ALUMINUM-BASE FUEL ELEMENTS  

SciTech Connect

Techniques were devised for quantitatively determining the accuracy of potentiometric uranium analyses in uranium-aluminum alloys containing up to 55 wt. % U and for evaluatin; the segregation existing in uraniumaluminum alloys containing as low as 7 wt. % U and as high as 50 wt. % U. A theory for predicting the mode of uranium segrcgation in these alloys was postulated. On the basis of the observed uranium segregation, the uranium content of a hypothetical fuel element was predicted by means of several sampling schemes. Dip sampling of the melt was demonstrated to be satisfactory for alloys containing 7 to 19 wt. % U. However, this technique was not considered suitable for alloys containiog 40 to 50 wt. % U, because a significant number of samples is required from the casting or the wrought alloy to adequately represent the fuel content. (auth) An accurate and rapid method for the volumetric determination of uranium has been developed for the ranges 0.3% to 90% uranium. The existing mercury cathode deposition equipment has been modified for the rapid removal of metallic impurities and for the electrolytic reduction of the uranium. (auth)

Thurber, W.C.; Beaver, R.J.

1958-09-19T23:59:59.000Z

348

Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium  

DOE Patents (OSTI)

A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material (low enriched U) circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

Wiencek, T.C.; Matos, J.E.; Hofman, G.L.

1993-12-31T23:59:59.000Z

349

Evaluation of a RF-Based Approach for Tracking UF6 Cylinders at a Uranium Enrichment Plant  

SciTech Connect

Approved industry-standard cylinders are used globally to handle and store uranium hexafluoride (UF{sub 6}) feed, product, tails, and samples at uranium enrichment plants. The International Atomic Energy Agency (IAEA) relies on time-consuming physical inspections to verify operator declarations and detect possible diversion of UF{sub 6}. Development of a reliable, automated, and tamper-resistant system for near real-time tracking and monitoring UF{sub 6} cylinders (as they move within an enrichment facility) would greatly improve the inspector function. This type of system can reduce the risk of false or misreported cylinder tare weights, diversion of nuclear material, concealment of excess production, utilization of undeclared cylinders, and misrepresentation of the cylinders contents. This paper will describe a proof-of-concept approach that was designed to evaluate the feasibility of using radio frequency (RF)-based technologies to track individual UF{sub 6} cylinders throughout a portion of their life cycle, and thus demonstrate the potential for improved domestic accountability of materials, and a more effective and efficient method for application of site-level IAEA safeguards. The evaluation system incorporates RF-based identification devices (RFID) which provide a foundation for establishing a reliable, automated, and near real-time tracking system that can be set up to utilize site-specific, rules-based detection algorithms. This paper will report results from a proof-of-concept demonstration at a real enrichment facility that is specifically designed to evaluate both the feasibility of using RF to track cylinders and the durability of the RF equipment to survive the rigors of operational processing and handling. The paper also discusses methods for securely attaching RF devices and describes how the technology can effectively be layered with other safeguard systems and approaches to build a robust system for detecting cylinder diversion. Additionally, concepts for off-site tracking of cylinders are described.

Pickett, Chris A [ORNL; Younkin, James R [ORNL; Kovacic, Donald N [ORNL; Laughter, Mark D [ORNL; Hines, Jairus B [ORNL; Boyer, Brian [Los Alamos National Laboratory (LANL); Martinez, B. [Los Alamos National Laboratory (LANL)

2008-01-01T23:59:59.000Z

350

FUEL ELEMENTS CONFERENCE, PARIS, NOVEMBER 18-23, 1957  

SciTech Connect

Papers are presented in the following major categories: applied metallurgical research, natural-uranium metallic fuel elements, enriched-uranium metallic fuel elements, nonmetallic fuel elements, corrosion of U alloys, irradiation effects on U, its alloys, and its compounds, and Pu fuel elements. (M.H.R.)

1958-10-31T23:59:59.000Z

351

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

SciTech Connect

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 {sup 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.

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

2003-06-13T23:59:59.000Z

352

A Multi-Country Analysis of Lifecycle Emissions From Transportation Fuels and Motor Vehicles  

E-Print Network (OSTI)

total world mine production of uranium, Australia producedin 2010 mine production will satisfy 75% of world uraniummine production worldwide, 0.50 for reprocessed tails and spent fuel, and 0.30 for military high-enriched uranium.

Delucchi, Mark

2005-01-01T23:59:59.000Z

353

A MULTI-COUNTRY ANALYSIS OF LIFECYCLE EMISSIONS FROM TRANSPORTATION FUELS AND MOTOR VEHICLES  

E-Print Network (OSTI)

total world mine production of uranium, Australia producedin 2010 mine production will satisfy 75% of world uraniummine production worldwide, 0.50 for reprocessed tails and spent fuel, and 0.30 for military high-enriched uranium.

Delucchi, Mark

2005-01-01T23:59:59.000Z

354

Application of the active well coincidence counter to the measurement of uranium  

Science Conference Proceedings (OSTI)

An Active Well Coincidence Counter has been developed to assay uranium fuel material in field inspection applications. The unit is used to measure bulk UO/sub 2/ samples, high enrichment uranium metals, LWR fuel pellets, and /sup 233/U-Th fuel materials which have very high gamma-ray backgrounds.

Menlove, H.O.; Foley, J.E.; Bosler, G.E.

1980-01-01T23:59:59.000Z

355

PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS  

DOE Patents (OSTI)

A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

Moore, R.H.

1962-10-01T23:59:59.000Z

356

Investigation of the low enrichment conversion of the Texas A and M Nuclear Science Center Reactor  

SciTech Connect

The use of highly enriched uranium as a fuel research reactors is of concern due to the possibility of diversion for nuclear weapons applications. The Texas A M TRIGA reactor currently uses 70% enriched uranium in a FLIP (Fuel Life Improvement Program) fuel element manufactured by General Atomics. Thus fuel also contains 1.5 weight percent of erbium as a burnable poison to prolong useful core life. US university reactors that use highly enriched uranium will be required to covert to 20% or less enrichment to satisfy Nuclear Regulatory Commission requirements for the next core loading if the fuel is available. This investigation examined the feasibility of a material alternate to uranium-zirconium hydride for LEU conversion of a TRIGA reactor. This material is a beryllium oxide uranium dioxide based fuel. The theoretical aspects of core physics analyses were examined to assess the potential advantages of the alternative fuel. A basic model was developed for the existing core configuration since it is desired to use the present fuel element grid for the replacement core. The computing approach was calibrated to the present core and then applied to a core of BeO-UO{sub 2} fuel elements. Further calculations were performed for the General Atomics TRIGA low-enriched uranium zirconium hydride fuel.

Reuscher, J.A.

1988-01-01T23:59:59.000Z

357

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

SciTech Connect

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.

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.

1985-02-01T23:59:59.000Z

358

Parametric Evaluation of Active Neutron Interrogation for the Detection of Shielded Highly-Enriched Uranium in the Field  

SciTech Connect

Parametric studies using numerical simulations are being performed to assess the performance capabilities and limits of active neutron interrogation for detecting shielded highly enriched uranium (HEU). Varying the shield material, HEU mass, HEU depth inside the shield, and interrogating neutron source energy, the simulations account for both neutron and photon emission signatures from the HEU with resolution in both energy and time. The results are processed to represent different irradiation timing schemes and several different classes of radiation detectors, and evaluated using a statistical approach considering signal intensity over background. This paper describes the details of the modeling campaign and some preliminary results, weighing the strengths of alternative measurement approaches for the different irradiation scenarios.

D. L. Chcihester; E. H. Seabury; S. J. Thompson; R. R. C. Clement

2011-10-01T23:59:59.000Z

359

Fuel  

E-Print Network (OSTI)

heavy-water-moderated, light-water-moderated and liquid-metal cooled fast breeder reactors fueled with natural or low-enriched uranium and containing thorium mixed with the uranium or in separate target channels. U-232 decays with a 69-year half-life through 1.9-year half-life Th-228 to Tl-208, which emits a 2.6 MeV gamma ray upon decay. We find that pressurized light-water-reactors fueled with LEU-thorium fuel at high burnup (70 MWd/kg) produce U-233 with U-232 contamination levels of about 0.4 percent. At this contamination level, a 5 kg sphere of U-233 would produce a gammaray dose rate of 13 and 38 rem/hr at 1 meter one and ten years after chemical purification respectively. The associated plutonium contains 7.5 percent of the undesirable heat-generating 88-year half-life isotope Pu-238. However, just as it is possible to produce weapon-grade plutonium in low-burnup fuel, it is also practical to use heavy-water reactors to produce U-233 containing only a few ppm of U-232 if the thorium is segregated in “target ” channels and discharged a few times more frequently than the natural-uranium “driver ” fuel. The dose rate from a 5-kg solid sphere of U-233 containing 5 ppm U-232 could be reduced by a further factor of 30, to about 2 mrem/hr, with a close-fitting lead sphere weighing about 100 kg. Thus the proliferation resistance of thorium fuel cycles depends very much upon how they are implemented. The original version of this manuscript was received by Science & Global Security on

Jungmin Kang A

2001-01-01T23:59:59.000Z

360

New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles  

E-Print Network (OSTI)

The comparison of nuclear facilities based on their barriers to nuclear material proliferation has remained a difficult endeavor, often requiring expert elicitation for each system under consideration. However, objectively comparing systems using a set of computable metrics to derive a single number representing a system is not, in essence, a nuclear nonproliferation specific problem and significant research has been performed for business models. For instance, Multi-Attribute Utility Analysis (MAUA) methods have been used previously to provide an objective insight of the barriers to proliferation. In this paper, the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR), a multi-tiered analysis tool based on the multiplicative MAUA method, is presented. It folds sixty three mostly independent metrics over three levels of detail to give an ultimate metric for nonproliferation performance comparison. In order to reduce analysts' bias, the weighting between the various metrics was obtained by surveying a total of thirty three nonproliferation specialists and nonspecialists from fields such as particle physics, international policy, and industrial engineering. The PRAETOR was used to evaluate the Fast Breeder Reactor Fuel Cycle (FBRFC). The results obtained using these weights are compared against a uniform weight approach. Results are presented for five nuclear material diversion scenarios: four examples include a diversion attempt on various components of a PUREX fast reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights, with and without safeguards in place. The numerical results corroborate nonproliferation truths and provide insight regarding fast reactor facilities' proliferation resistance in relation to known standards.

Metcalf, Richard R.

2009-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

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

SciTech Connect

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

Matos, J.E.; Freese, K.E.

1986-11-03T23:59:59.000Z

362

Nuclear Fuels  

Science Conference Proceedings (OSTI)

Mar 13, 2012... diffusion couple studies using cerium as a surrogate for uranium, ... reactors from highly enriched (HEU) to low enriched (LEU) uranium, ...

363

Power Surge: Uranium alloy fuel for TerraPower | Y-12 National Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Power Surge: Uranium alloy ... Power Surge: Uranium alloy ... Power Surge: Uranium alloy fuel for TerraPower Posted: July 18, 2012 - 9:45am | Y-12 Report | Volume 9, Issue 1 | 2012 Since 2010, Y-12 has provided TerraPower with technical support in the fabrication methods for uranium alloy fuel to be used in a new traveling wave nuclear reactor that can run for more than 30 years without refueling. Image of reactor power concept, used with permission of TerraPower, LLC. Y-12's nuclear expertise, expanding since the site's integral role in the Manhattan Project, is positioning the Y-12 Complex at the forefront of what Sen. Lamar Alexander repeatedly asserts is needed - "a new Manhattan Project for clean energy independence." TerraPower, a private company backed by Microsoft founder Bill Gates, is

364

Thermal analysis of uranium zirconium hydride fuel using a lead-bismuth gap at LWR operating temperatures  

E-Print Network (OSTI)

Next generation nuclear technology calls for more advanced fuels to maximize the effectiveness of new designs. A fuel currently being studied for use in advanced light water reactors (LWRs) is uranium zirconium hydride ...

Ensor, Brendan M. (Brendan Melvin)

2012-01-01T23:59:59.000Z

365

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

SciTech Connect

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

1993-07-01T23:59:59.000Z

366

Benchmark calculations for the diluted highly enriched uranium (HEU) and aluminum experiment  

SciTech Connect

The HEU-Al experiment was performed using the Planet universal critical assembly at Los Alamos Critical Experiment Facility (LACEF) in Los Alamos National Laboratory. This experiment consisted of placing HEU foils interspersed with aluminum plates in a column stack. These uranium foils were moderated and reflected by polyethylene square plates. This experiment was performed to measure the prompt neutron decay constants in uranium systems diluted by matrix materials. This experimental set-up yielded a Al/235U ratio of 60:1.1 The experimental keff was 1.001 and the modeled MCNP keff was 1.0016{+-}0.0004. This report summarizes the benchmark calculations performed to validate the experiment. The experimental arrangement is depicted in Figure 1. As Figure 1 illustrates the stack is divided into two parts. The bottom half of the stack rest on an aluminum support plate which is 1 inch thick. The top half of the experiment rest on 0.75 inch thick polyethylene plate. Criticality is achieved by decreasing the gap between the top and bottom portions of the stack. To disassemble the configuration the bottom stack is dropped to its initial position. There are no other control or safety rods inside the assembly.

Loaiza, D. J. (David J.); Sanchez, R. G. (Rene G.)

2001-01-01T23:59:59.000Z

367

Beyond Basic Target Enrichment: New Tools to Fuel Your NGS Research ( 7th Annual SFAF Meeting, 2012)  

Science Conference Proceedings (OSTI)

Jennifer Carter on "Beyond Basic Target Enrichment: New Tools to fuel your NGS Research" at the 2012 Sequencing, Finishing, Analysis in the Future Meeting held June 5-7, 2012 in Santa Fe, New Mexico.

Carter, Jennifer [Agilent

2012-06-01T23:59:59.000Z

368

DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS  

DOE Green Energy (OSTI)

The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl/sub 4/ or UAl/sub 3/ compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm/sup 3/), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze the plates into fuel assemblies. Corrosion tests of the fuel material were conducted in 20 and 60 deg C water to determine the integrity of the fuel material in the event of sin inadventent cladding failure. In addition, specimens were prepared to evaluate penformance under extensive irradiation Prior to studying the uranium carbide-aluminum system, methods for preparing the carbides were investigated. Are melting uranium and carnon was satisfactory for obtaining small quantities of various carbides. Later, reaction of graphite with UO/sub 2/ was successfully employed in the preparation of large quantities of UC/sub 2/, Studies of the chemical compatibility of cold-pressed compacts containing 50 wt% uranium carbide dispersed in aluminum revealed a marked trend toward stebifity as the carbon content of the uranium carbide increased from 446 to 9.20% C. Severe volume increases occurred in monocarbide dispersions with attendant formation of large quantities of the uranium-allumnim inter-metallic compounds. Dicarbide dispersions, on the other band, exhibited negligible reaction with aluminum after extended periods at 580 and 620 deg C. However, it was demonstrated that hydrogen can promote a reaction in UC/sub 2/-Al compacts. The hydrogen appears to reduce the UC/sub 2/ to UC which can subsequently react with aluminum producing the previously noted deleterious effects. A growth study at 605 deg C of composite fuel plates containing 59 wt.% UC/sub 2/ revealed insignificant changes within processing periods envisioned for fuel element processing. However, plate elongations as high as 2.5% were observed after 100 hr at this temperature. Severe blistering which occurred on fuel plates fabricated in the initial stages of the investigation was attributed to gaseous hydrocarbons, and the condition was ellminated by vacuum degasification of cold-pressed compacts. With the exception of the degasification requirement, procedures for manufacturing UC- bearing fuel elements were identical to those specified for the Geneva Conference Reactor fuel elements. Dispersions of uranium dicarbide corroded catastrophically in 20 and 60 deg C water, thus limiting the application of this material However, spocimens were prepared and insented in the MTR to evaluate the irradiation behavior of this fuel because of its potential application in onganic- cooled reactors. (auth)

Thurber, W.C.; Beaver, R.J.

1959-11-19T23:59:59.000Z

369

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies  

SciTech Connect

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

Chodak, P. III

1996-05-01T23:59:59.000Z

370

The central void reactivity in the Oak Ridge enriched uranium (93.2) metal sphere  

SciTech Connect

The central reactivity void worth was measured in the Oak Ridge unmoderated and unreflected uranium (93.20 wt% {sup 235}U) metal sphere by replacement measurements in a small (0.460-cm-diam) central spherical region in an 8.7427-cm-radius sphere. The central void worth was 9.165 {+-} 0.023 cents using the delayed neutron relative abundances and decay constants of Keepin, Wimett, and Zeigler to obtain the reactivity in cents from the stable reactor period measurements using the Inhour equation. This value is slightly larger than measurements with GODIVA 1 with larger cylindrical samples of uranium (93.70 wt% {sup 235}U) in the center: 135.50 {+-} 0.12 cents/mole for GODIVA 1 and 138.05 {+-} 0.34 cents/mole for the Oak Ridge sphere measurements, and the difference could be due to sample size effect. The central worth in {Delta}k units was calculated by neutron transport theory methods to be 6.02 {+-} 0.01 x 10{sup {minus}4} {Delta}k. The measured and calculated values are related by the effective delayed neutron fraction. The value of the effective delayed neutron fraction obtained in this way from the Oak Ridge sphere is 0.00657 {+-} 0.00002, which is in excellent agreement with that obtained from GODIVA 1 measurements, where the effective delayed neutron fraction was determined as the increment between delayed and prompt criticality and was 0.0066. From these Oak Ridge measurements, using the delayed neutron parameters of ENDF-B/VI to obtain the reactivity from the stable reactor period measurements, the central void worth is 7.984 {+-} 0.021 cents, and the inferred effective delayed neutron fraction is 0.00754. This central void worth and effective delayed neutron fractions are 14.2% higher than those obtained from use of the Keepin et al. delayed neutron data and produce a value of delayed neutron fraction in disagreement with GODIVA 1 measurements, thus questioning the usefulness of the relative abundances and decay constants of the six-group delayed neutron parameters of ENDF-B/VI for uranium for obtaining the reactivity from the measured reactor period using the Inhour equation.

Milhalczo, J.T.; Lynn, J.J.; Taylor, J.R.

1997-03-01T23:59:59.000Z

371

Parametric Study of Front-End Nuclear Fuel Cycle Costs  

Science Conference Proceedings (OSTI)

This study provides an overview of front-end fuel cost components for nuclear plants, specifically uranium concentrates, uranium conversion services, uranium enrichment services, and nuclear fuel fabrication services. A parametric analysis of light-water reactor (LWR) fuel cycle costs is also included in order to quantify the impacts that result from changes in the cost of one or more front-end components on overall fuel cycle costs.

2009-02-20T23:59:59.000Z

372

NEUTRALIZATIONS OF HIGH ALUMINUM LOW URANIUM USED NUCLEAR FUEL SOLUTIONS CONTAINING GADOLINIUM AS A NEUTRON POISON  

DOE Green Energy (OSTI)

H-Canyon will begin dissolving High Aluminum - Low Uranium (High Al/Low U) Used Nuclear Fuel (UNF) following approval by DOE which is anticipated in CY2011. High Al/Low U is an aluminum/enriched uranium UNF with small quantities of uranium relative to aluminum. The maximum enrichment level expected is 93% {sup 235}U. The High Al/Low U UNF will be dissolved in H-Canyon in a nitric acid/mercury/gadolinium solution. The resulting solution will be neutralized and transferred to Tank 39H in the Tank Farm. To confirm that the solution generated could be poisoned with Gd, neutralized, and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of simulant solutions was examined. Experiments were performed with three simulant solutions representative of the H-Canyon estimated concentrations in the final solutions after dissolution. The maximum U, Gd, and Al concentration were selected for testing from the range of solution compositions provided. Simulants were prepared in three different nitric acid concentrations, ranging from 0.5 to 1.5 M. The simulant solutions were neutralized to four different endpoints: (1) just before a solid phase was formed (pH 3.5-4), (2) the point where a solid phase was obtained, (3) 0.8 M free hydroxide, and (4) 1.2 M free hydroxide, using 50 wt % sodium hydroxide (NaOH). The settling behavior of the neutralized solutions was found to be slower compared to previous studies, with settling continuing over a one week period. Due to the high concentration of Al in these solutions, precipitation of solids was observed immediately upon addition of NaOH. Precipitation continued as additional NaOH was added, reaching a point where the mixture becomes almost completely solid due to the large amount of precipitate. As additional NaOH was added, some of the precipitate began to redissolve, and the solutions neutralized to the final two endpoints mixed easily and had expected densities of typical neutralized waste. Based on particle size and scanning electron microscopy analyses, the neutralized solids were found to be homogeneous and less than 20 microns in size. The majority of solids were less than 4 microns in size. Compared to previous studies, a larger percentage of the Gd was found to precipitate in the partially neutralized solutions (at pH 3.5-4). In addition the Gd:U mass ratio was found to be at least 1.0 in all of the solids obtained after partial or full neutralization. The hydrogen to U (H:U) molar ratios for two accident scenarios were also determined. The first was for transient neutralization and agitator failure. Experimentally this scenario was determined by measuring the H:U ratio of the settled solids. The minimum H:U molar ratio for solids from fully neutralized solutions was 388:1. The second accident scenario is for the solids drying out in an unagitiated pump box. Experimentally, this scenario was determined by measuring the H:U molar ratio in centrifuged solids. The minimum H:U atom ratios for centrifuged precipitated solids was 250:1. It was determined previously that a 30:1 H:Pu atom ratio was sufficient for a 1:1 Gd:Pu mass ratio. Assuming a 1:1 equivalence with {sup 239}Pu, the results of these experiments show Gd is a viable poison for neutralizing U/Gd solutions with the tested compositions.

Taylor-Pashow, K.

2011-06-08T23:59:59.000Z

373

Uranium Oxide Semiconductors  

NLE Websites -- All DOE Office Websites (Extended Search)

of semiconductors, it would consume the annual production rate of depleted uranium from uranium enrichment facilities. For more information: PDF Semiconductive Properties of...

374

L'URANIUM ET LES ARMES L'URANIUM APPAUVRI. Pierre Roussel*  

E-Print Network (OSTI)

(depleted uranium) · 4 oxidation states (+4, +6 most common) · U(VI) water-soluble, U(IV) in-soluble Metals Uranium ­ heaviest natural element - 17 isotopes · Natural form % = U-238 (99.27), U-235 (0.72), U-234 (0 in nuclear fuel ­ U-235 (readily fissionable) · Used in nuclear and conventional weapons · Uranium enrichment

Paris-Sud XI, Université de

375

Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond  

SciTech Connect

The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M. [Candu Energy Inc., 2285 Speakman Drive, Mississauga, ON L5K 1B1 (Canada)

2012-07-01T23:59:59.000Z

376

Direct fissile assay of enriched uranium using random self-interrogation and neutron coincidence response  

DOE Patents (OSTI)

Apparatus and method for the direct, nondestructive evaluation of the .sup.235 U nuclide content of samples containing UF.sub.6, UF.sub.4, or UO.sub.2 utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1.sigma.) for cylinders containing UF.sub.6 with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF.sub.6 takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures.

Menlove, Howard O. (Los Alamos, NM); Stewart, James E. (Los Alamos, NM)

1986-01-01T23:59:59.000Z

377

Direct fissile assay of enriched uranium using random self-interrogation and neutron coincidence response  

DOE Patents (OSTI)

Apparatus and method for the direct, nondestructive evaluation of the /sup 235/U nuclide content of samples containing UF/sub 6/, UF/sub 4/, or UO/sub 2/ utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1sigma) for cylinders containing UF/sub 6/ with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF/sub 6/ takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures. 4 figs., 1 tab.

Menlove, H.O.; Stewart, J.E.

1985-02-04T23:59:59.000Z

378

Simultaneous measurements of plutonium and uranium in spent-fuel dissolver solutions  

Science Conference Proceedings (OSTI)

The authors have studied the isotope dilution gamma-ray spectrometry (IDGS) technique for simultaneous measurements of elemental concentrations and isotopic compositions for both plutonium and uranium in input spent-fuel dissolver solutions at a reprocessing plant. The technique under development includes both sample preparation and analysis methods. For simultaneous measurements of both plutonium and uranium, a critical issue is to develop a new method to keep both plutonium and uranium in the sample after they are separated from fission products. Furthermore, it is equally important to improve the analysis method so that the precision and accuracy of the plutonium analysis remain unaffected while uranium is retained in the sample. To keep both plutonium and uranium in the sample for simultaneous measurements, extraction chromatography is being studied and shows promise to achieve the goal of cosegregation of the plutonium and uranium. The technique uses U/TEVA{center_dot}Spec resin to separate fission products and recover both uranium and plutonium in the resin from dissolver solutions for subsequent measuring using high-resolution gamma-ray spectrometry. Owing to the fact that the U/Pu ratio is altered during the fission product separation phase, it is necessary to develop a method which could accurately correct for this effect. Such a method was developed using the unique decay properties of {sup 241}Pu to {sup 237}U and shows considerable promise in allowing for accurate determination of the {sup 235}U concentrations before the chemical extraction.

Li, T.K. [Los Alamos National Lab., NM (United States); Kuno, T.; Kitagawa, O.; Sato, S.; Kurosawa, A.; Kuno, Y. [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan)

1997-11-01T23:59:59.000Z

379

Neutronics studies of uranium-based fully ceramic micro-encapsulated fuel for PWRs  

Science Conference Proceedings (OSTI)

This study evaluates the core neutronics and fuel cycle characteristics using uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR assembly designs with FCM fuel have been developed, which by virtue of their TRISO particle-based elements are expected to achieve higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software used to model the assembly designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation (RPT) method was used for lattice calculations due to the long run times associated with the SCALE DH capability. In order to understand the impact on reactivity and reactor operating cycle length, a parametric study was performed by varying TRISO particle design features, such as kernel diameter, coating layer thicknesses, and packing fraction. Also, other features such as the selection of matrix material (SiC, zirconium) and fuel rod dimensions were studied. After evaluating different uranium-based fuels, the higher compound density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime and temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. (authors)

George, N. M.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States); Terrani, K.; Godfrey, A.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2012-07-01T23:59:59.000Z

380

Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor  

Science Conference Proceedings (OSTI)

A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

1980-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel enrichment" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

A System Dynamics Study of Uranium and the Nuclear Fuel Cycle  

E-Print Network (OSTI)

market modelling’. Energy Economics, Vol. 33, Issue 5, pp. 840–852. Kiani, B., Mirzamohammadi, S., Hosseini, S.H. (2010). ‘A Survey on the Role of System Dynamics Methodology on Fossil Fuel Resources Analysis’. International Business Research, Vol... A System Dynamics Study of Uranium and the Nuclear Fuel Cycle Matthew Rooney, William J. Nuttall and Nikolas Kazantzis May 2013 CWPE 1319 & EPRG 1311 www.eprg.group.cam.ac.uk A System Dynamics...

Rooney, Matthew; Nuttall, William J.; Kazantzis, Nikolas

2013-05-31T23:59:59.000Z

382

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report  

SciTech Connect

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Dis

Sean M. McDeavitt

2011-04-29T23:59:59.000Z

383

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

SciTech Connect

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

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

2012-07-01T23:59:59.000Z

384

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

SciTech Connect

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.

1998-08-21T23:59:59.000Z

385

Preliminary investigations for technology assessment of /sup 99/Mo production from LEU (low enriched uranium) targets. [For production of /sup 99m/Tc; by different methods  

SciTech Connect

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 /sup 99/Mo. Issues that were addressed are: (1) purity and yield of the /sup 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 /sup 99/Mo production. 37 refs., 1 fig., 5 tabs.

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

1986-11-01T23:59:59.000Z

386

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

without need for uranium enrichment capability or actinidesfor expanding the uranium enrichment capacity and withoutleft over from the uranium enrichment process: (1) Per unit

Heidet, Florent

2010-01-01T23:59:59.000Z

387

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

SciTech Connect

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.

1995-12-01T23:59:59.000Z

388

The Fuel Fabrication Capability and Uranium-molybdenum Alloy  

Science Conference Proceedings (OSTI)

Abstract Scope, The Fuel Fabrication Capability (FFC) is part of the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) Global ...

389

DOE/EA-1471: Environmental Assessment for the Transportation of Highly Enriched Uranium from the Russian Federation to the Y-12 National Security Complex and Finding of No Significant Impact (January 2004)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EA for the Transportation of Highly Enriched Uranium from the Russian Federation to the Y-12 National Security Complex EA for the Transportation of Highly Enriched Uranium from the Russian Federation to the Y-12 National Security Complex i FINDING OF NO SIGNIFICANT IMPACT FOR THE TRANSPORTATION OF HIGHLY ENRICHED URANIUM FROM THE RUSSIAN FEDERATION TO THE Y-12 NATIONAL SECURITY COMPLEX ISSUED BY: United States Department of Energy ACTION: Finding of No Significant Impact SUMMARY: The United States (U.S.) Department of Energy (DOE) proposes to transport highly enriched uranium (HEU) from Russia to a secure storage facility in Oak Ridge, TN. This proposed action would allow the United States and Russia to accelerate the disposition of excess nuclear weapons materials in the interest of promoting nuclear disarmament, strengthening nonproliferation, and combating terrorism. The HEU

390

Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed  

Science Conference Proceedings (OSTI)

A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

Smirnov, A. Yu., E-mail: a.y.smirnoff@rambler.ru; Sulaberidze, G. A. [National Research Nuclear University MEPhI (Russian Federation); Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A., E-mail: neva@dhtp.kiae.ru; Proselkov, V. N.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

2012-12-15T23:59:59.000Z

391

NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs  

Science Conference Proceedings (OSTI)

This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. Key Words: FCM, TRISO, Uranium Mononitride, PWR

George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

2012-01-01T23:59:59.000Z

392

Domestic Uranium Production Report 2004 -2011  

U.S. Energy Information Administration (EIA)

Nuclear & Uranium. Uranium fuel, nuclear reactors, generation, spent fuel. Total Energy. Comprehensive data summaries, comparisons, analysis, and projections ...

393

PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle  

Science Conference Proceedings (OSTI)

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

Bi, G.; Liu, C.; Si, S. [Shanghai Nuclear Engineering Research and Design Inst., No. 29, Hongcao Road, Shanghai, 200233 (China)

2012-07-01T23:59:59.000Z

394

Y-12 Knows Uranium | Y-12 National Security Complex  

NLE Websites -- All DOE Office Websites (Extended Search)

Knows Uranium Knows Uranium Y-12 Knows Uranium Posted: July 22, 2013 - 3:45pm | Y-12 Report | Volume 10, Issue 1 | 2013 Y-12 produces many forms of uranium. They may be used in chemical processing steps on-site or shipped elsewhere to serve as raw materials for nuclear fuel or as research tools. All of uranium's uses, defense related and otherwise, are critical to the nation. Y-12's understanding of uranium, coupled with the site's work with enriched uranium metal, alloys, oxides, compounds and solutions, is unique in the Nuclear Security Enterprise. "The Y-12 work force understands both established uranium science and the esoteric things related to uranium's behavior," said engineer Alan Moore. "Such a deep, detailed understanding comes from experience,

395

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel  

NLE Websites -- All DOE Office Websites (Extended Search)

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Vitaly T. Gotovchikov a , Victor A. Seredenko a , Valentin V. Shatalov a , Vladimir N. Kaplenkov a , Alexander S. Shulgin a , Vladimir K. Saranchin a , Michail A. Borik a∗ , Charles W. Forsberg b , All-Russian Research Institute of Chemical Technology (ARRICT) 33, Kashirskoe ave., Moscow, Russia, 115409, E-mail: chem.conv@ru.net Oak Ridge National Laboratory (ORNL) Bethel Wall Road, P.O. Box 2008, MS-6165, Oak Ridge, TN, USA, 37831 Abstract - Induction cold crucible melters (ICCM) have the potential to be a very-low-cost high-throughput method for the production of DUO 2 for SNF casks. The proposed work would develop these melters for this specific application. If a

396

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors  

DOE Patents (OSTI)

A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

McLean, W. II; Miller, P.E.

1997-12-16T23:59:59.000Z

397

Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes  

SciTech Connect

A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

Forsberg, C.W.

1997-03-01T23:59:59.000Z

398

Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials  

SciTech Connect

A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties.

Forsberg, C.W.; Haire, M.J.

1997-07-07T23:59:59.000Z

399

CHARACTERIZATION OF SURFACTANTS IN ALUMINUM-URANIUM FUEL REPROCESSING SOLUTIONS  

SciTech Connect

Surface active materials in aluminum nitrate-nitric acid fuel reprocessing solutions were characterized. Polymerized silica, zirconium- modified silica and soluble dibutyl phosphate species were found to contribute to stable emulsion formation. These surfactants were reduced in effectiveness by added acid. (auth)

Cannon, R.D.

1959-10-20T23:59:59.000Z

400

Uranium resource utilization improvements in the once-through PWR fuel cycle  

Science Conference Proceedings (OSTI)

In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U/sub 3/O/sub 8/ consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout.

Matzie, R A [ed.

1980-04-01T23:59:59.000Z

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401

Disposition of DOE Excess Depleted Uranium, Natural Uranium, and  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium The U.S. Department of Energy (DOE) owns and manages an inventory of depleted uranium (DU), natural uranium (NU), and low-enriched uranium (LEU) that is currently stored in large cylinders as depleted uranium hexafluoride (DUF6), natural uranium hexafluoride (NUF6), and low-enriched uranium hexafluoride (LEUF6) at the DOE Paducah site in western Kentucky (DOE Paducah) and the DOE Portsmouth site near Piketon in south-central Ohio (DOE Portsmouth)1. This inventory exceeds DOE's current and projected energy and defense program needs. On March 11, 2008, the Secretary of Energy issued a policy statement (the

402

United States Department of Energy, Office of Environmental Management, Uranium Enrichment Decontamination and Decomissioning Fund financial statements, September 30, 1996 and 1995  

SciTech Connect

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.

1997-05-01T23:59:59.000Z

403

Advanced Neutron Source enrichment study -- Volume 1: Main report. Final report, Revision 12/94  

SciTech Connect

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

Bari, R.A.; Ludewig, H.; Weeks, J. [Brookhaven National Lab., Upton, NY (United States). Dept. of Advanced Technology

1994-12-31T23:59:59.000Z